ML17355A661

Amendment 64 to Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant

ML17355A661
Person / Time
Site:	Columbia
Issue date:	12/31/2017
From:	Energy Northwest
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17355A655	List: GO2-17-190, Amendment 64 to Final Safety Analysis Report, Chapter 12, Radiation Protection ML17355A654 ML17355A656 ML17355A657 ML17355A658 ML17355A660 ML17355A661 ML17355A662 ML17355A663 ML17355A664 ML17355A666 ML17355A667 ML17355A668 ML17355A670
References
GO2-17-190
Download: ML17355A661 (603)

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C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

TABLE OF CONTENTS

Section Page 1-i

1.1 INTRODUCTION

.........................................................................1.1-1

1.2 GENERAL

PLANT DESCRIPTION...................................................1.2-1 1.2.1 PRINCIPAL DESIGN CRITERIA..................................................1.2-1 1.2.1.1 General Design Criteria............................................................1.2-1 1.2.1.1.1 Power Generation Design Criteria.............................................1.2-1 1.2.1.1.2 Safety Design Criteria............................................................1.2-2 1.2.1.2 System Criteria......................................................................1.2-5 1.2.1.2.1 Nuclear System Criteria..........................................................1.2-5 1.2.1.2.2 Power Conversion System Criteria............................................1.2-6 1.2.1.2.3 Electrical Power Systems Criteria..............................................1.2-6 1.2.1.2.4 Radwaste System Criteria........................................................1.2-7 1.2.1.2.5 Auxiliary Systems Criteria......................................................1.2-7 1.2.1.2.6 Shielding and Access Co ntrol Criteria.........................................1.2-7 1.2.1.2.7 Nuclear Safety Systems and Engineered Safety Features Criteria........1.2-8 1.2.1.2.8 Process Control Systems Criteria..............................................1.2-8 1.2.1.3 Plant Design Criteria................................................................1.2-9 1.2.2 PLANT DESCRIPTION..............................................................1.2-10 1.2.2.1 Site Characteristics..................................................................1.2-10 1.2.2.1.1 Site Location and Size............................................................1.2-10 1.2.2.1.2 Description of Site Environs....................................................1.2-10 1.2.2.1.2.1 Site Land.........................................................................1.2-10 1.2.2.1.2.2 Population........................................................................1.2-10 1.2.2.1.2.3 Land Use.........................................................................1.2-10 1.2.2.1.2.4 Meteorology.....................................................................1.2-10 1.2.2.1.2.5 Hydrology........................................................................1.2-10 1.2.2.1.2.6 Geology...........................................................................1.2-10 1.2.2.1.2.7 Seismology.......................................................................1.2-11 1.2.2.1.3 Design Basis Depending on Site Environs....................................1.2-11 1.2.2.2 General Arrangement of Structures and Equipment...........................1.2-12 1.2.2.3 Symbols Used on Engineering Drawings........................................1.2-13 1.2.2.4 Nuclear System......................................................................1.2-13 1.2.2.4.1 Reactor Core and Control Rods................................................1.2-13 1.2.2.4.2 Reactor Vessel and Internals....................................................1.2-13 1.2.2.4.3 Reactor Recirculation System...................................................1.2-14 1.2.2.4.4 Residual Heat Removal System.................................................1.2-14 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

TABLE OF CONTENTS

Section Page LDCN-13-008 1-ii 1.2.2.4.5 Reactor Water Cleanup System ................................................. 1.2-15 1.2.2.4.6 Nuclear Leak Detection System ................................................ 1.2-15 1.2.2.5 Nuclear Safety Systems and Engineered Safety Features

..................... 1.2-15 1.2.2.5.1 Reactor Protection System ....................................................... 1.2-15 1.2.2.5.2 Neutron Monitoring System ..................................................... 1.2-16 1.2.2.5.3 Control Rod Drive System ...................................................... 1.2-16 1.2.2.5.4 Control Rod Drive Housing Supports ......................................... 1.2-16 1.2.2.5.5 Control Rod Velocity Limiter ................................................... 1.2-16 1.2.2.5.6 Pressure Relief System (Nuclear System)

..................................... 1.2-16 1.2.2.5.7 Reactor Core Isolation Cooling System ....................................... 1.2-17 1.2.2.5.8 Emergency Core Cooling System .............................................. 1.2-17 1.2.2.5.8.1 High-Pressure Core Spray System ........................................... 1.2-17 1.2.2.5.8.2 Automatic Depressurization System ......................................... 1.2-17 1.2.2.5.8.3 Low-Pressure Core Spray System ........................................... 1.2-17 1.2.2.5.8.4 Low-Pressure Coolant Injection .............................................. 1.2-18 1.2.2.5.9 Primary Containment ............................................................. 1.2-18 1.2.2.5.9.1 Functional Design

............................................................... 1.2-18 1.2.2.5.9.2 Drywell Cooling System ....................................................... 1.2-18 1.2.2.5.9.3 Suppression Pool Cooling ..................................................... 1.2-19 1.2.2.5.9.4 Containment Spray.............................................................. 1.2-19 1.2.2.5.9.5 Containment Atmosphere Control ........................................... 1.2-19 1.2.2.5.10 Primary Containment and Reactor Vessel Isolation System. .............. 1.2-19 1.2.2.5.11 Main Steam Line Isolation Valves ............................................. 1.2-20 1.2.2.5.12 Main Steam Line Flow Restrictors

............................................. 1.2-20 1.2.2.5.13 Main Steam Line Radiation Monitoring System ............................. 1.2-20 1.2.2.5.14 Standby Service Water and High-Pressure Core Spray Service Water Systems ..................................................................... 1.2-20 1.2.2.5.15 Reactor Building - Secondary Containment .................................. 1.2-21 1.2.2.5.16 Reactor Building Ventilation Exhaust Radiation Monitoring System .... 1.2-21 1.2.2.5.17 Standby Gas Treatment System ................................................. 1.2-22 1.2.2.5.18 Standby Alternating Current Power Supply System

......................... 1.2-22 1.2.2.5.19 Direct Current Power Supply System ......................................... 1.2-23 1.2.2.5.20 Standby Liquid Control System

................................................. 1.2-23 1.2.2.5.21 Safe Shutdown from Outside the Main Control Room ..................... 1.2-23 1.2.2.5.22 Main Steam Line Isolation Valve Leakage Control System (Deactivated) ....................................................................... 1.2-24 C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

TABLE OF CONTENTS

Section Page LDCN-15-033 1-iii 1.2.2.5.23 Fuel Pool Cooling and Cleanup System ....................................... 1.2-24 1.2.2.5.24 Hardened Containment Vent (HCV) System .............................................1.2-24 1.2.2.6 Power Conversion System ......................................................... 1.2-24 1.2.2.6.1 Turbine Ge nerator................................................................. 1.2-24 1.2.2.6.2 Main Steam System ............................................................... 1.2-25 1.2.2.6.3 Main Condenser ................................................................... 1.2-25 1.2.2.6.4 Main Condenser Evacuation System ........................................... 1.2-25 1.2.2.6.5 Turbine Gland Seal System ...................................................... 1.2-25 1.2.2.6.6 Steam Bypass System and Pressure Control System ........................ 1.2-26 1.2.2.6.7 Circulating Water System

........................................................ 1.2-26 1.2.2.6.8 Condensate and Feedwater System

............................................. 1.2-26 1.2.2.6.9 Condensate Filter-Demineralizer System ..................................... 1.2-26 1.2.2.7 Electrical Systems, Instrumentation, and Control ............................. 1.2-26 1.2.2.7.1 Electrical Power Systems ........................................................ 1.2-26 1.2.2.7.2 Electrical Power Systems Process Control and Instrumentation .......... 1.2-27 1.2.2.7.3 Nuclear System Process Control and Instrumentation ...................... 1.2-27 1.2.2.7.3.1 Reactor Manual Control System ............................................. 1.2-27 1.2.2.7.3.2 Recirculation Flow Control System

.......................................... 1.2-27 1.2.2.7.3.3 Neutron Monitoring System

................................................... 1.2-28 1.2.2.7.3.4 Refueling Interlocks ............................................................ 1.2-28 1.2.2.7.3.5 Reactor Vessel Instrumentation

............................................... 1.2-28 1.2.2.7.3.6 Process Computer System ..................................................... 1.2-28 1.2.2.7.4 Power Conversion Systems Process Control and Instrumentation ........ 1.2-28 1.2.2.7.4.1 Digital Electro-Hydraulic Control System .................................. 1.2-28 1.2.2.7.4.2 Feedwater System Control .................................................... 1.2-29 1.2.2.8 Radioactive Waste Systems ........................................................ 1.2-29 1.2.2.8.1 Liquid Radwaste System ......................................................... 1.2-29 1.2.2.8.2 Solid Radwaste System ........................................................... 1.2-29 1.2.2.8.3 Gaseous Radwaste System ....................................................... 1.2-30 1.2.2.9 Radiation Monitoring and Control

................................................ 1.2-30 1.2.2.9.1 Process Radiation Monitoring

................................................... 1.2-30 1.2.2.9.2 Area Radiation Monitors ......................................................... 1.2-31 1.2.2.9.3 Site Radiological Environmental Monitoring ................................. 1.2-31 1.2.2.9.4 Liquid Radwaste System Control ............................................... 1.2-31 1.2.2.9.5 Solid Radwaste System Control ................................................ 1.2-31 1.2.2.9.6 Gaseous Radwaste System Control

............................................. 1.2-31 C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

TABLE OF CONTENTS (Continued)

Section Page 1-iv 1.2.2.10 Shielding .............................................................................. 1.2-32 1.2.2.11 Fuel Handling and Storage Systems .............................................. 1.2-32 1.2.2.11.1 New and Spent Fuel Storage .................................................... 1.2-32 1.2.2.11.2 Fuel Handling System ............................................................ 1.2-32 1.2.2.11.3 Fuel Pool Cooling and Cleanup System ....................................... 1.2-32 1.2.2.12 Cooling Water and Auxiliary Systems ........................................... 1.2-33 1.2.2.12.1 Reactor Building Closed Cooling Water System ............................ 1.2-33 1.2.2.12.2 Plant Service Water System ..................................................... 1.2-33 1.2.2.12.3 Ultimate Heat Sink ................................................................ 1.2-33 1.2.2.12.4 Demineralized Water Makeup System

......................................... 1.2-33 1.2.2.12.5 Potable Water and Sanitary Drain Systems ................................... 1.2-33 1.2.2.12.6 Process Sampling Systems ....................................................... 1.2-33 1.2.2.12.7 Condensate Supply System ...................................................... 1.2-34 1.2.2.12.8 Equipment and Floor Drainage Systems ...................................... 1.2-34 1.2.2.12.9 Compressed Air Systems ........................................................ 1.2-34 1.2.2.12.10 Heating, Ventilating, and Air Conditioning Systems ....................... 1.2-35 1.2.2.12.11 Fire Protection System ........................................................... 1.2-36 1.2.2.12.12 Communications Systems ........................................................ 1.2-37 1.2.2.12.13 Lighting Systems

.................................................................. 1.2-37 1.2.2.12.14 Normal Auxiliary Alternating Current Power System ...................... 1.2-37 1.2.2.12.15 Diesel Generator Fuel Oil Storage and Transfer System ................... 1.2-38 1.2.2.12.16 Auxiliary Steam System .......................................................... 1.2-38 1.2.3 COMPLIANCE WITH NRC REGULATORY GUIDES ....................... 1.2-38

1.3 COMPARISON

TABLES ................................................................ 1.3-1 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS ..................... 1.3-1 1.3.1.1 Nuclear Steam Supply System Design Characteristics ........................ 1.3-1 1.3.1.2 Power Conversion System Design Characteristics

............................. 1.3-1 1.3.1.3 Engineered Safety Features Design Characteristics ........................... 1.3-1 1.3.1.4 Containment Design Characteristics ............................................. 1.3-1 1.3.1.5 Radioactive Waste Management Systems Design Characteristics ........... 1.3-1 1.3.1.6 Structural Design Characteristics ................................................. 1.3-1 1.3.1.7 Electrical Power Systems Design Characteristics .............................. 1.3-1 1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION

....... 1.3-2 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

TABLE OF CONTENTS (Continued)

Section Page LDCN-15-011 1-v 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS ....................... 1.4-1 1.4.1 APPLICANT/OPERATOR

........................................................... 1.4-1 1.4.2 ENGINEER AND CONSTRUCTION MANAGEMENT -

BURNS & ROE, INC. ................................................................. 1.4-1

1.4.3 NUCLEAR

STEAM SYSTEM SUPPLIER - GENERAL ELECTRIC COMPANY .............................................................................. 1.4-1

1.4.4 TURBINE

GENERATOR SUPPLIER - WESTINGHOUSE ELECTRIC

CORPORATION.

...................................................................... 1.4-2 1.4.5 SYSTEM COMPLETION CONTRACTOR - BECHTEL ...................... 1.4-2

1.4.6 MAJOR

CONTRACTORS

............................................................ 1.4-2 1.4.6.1 Fischbach/Lord

...................................................................... 1.4-2 1.4.6.2 Pittsburgh-Des Moines Steel Company .......................................... 1.4-3 1.4.6.3 Wright - Schuchart - Harbor/Boecon (Boeing Construction) General Energy Resources, Inc. .................................................. 1.4-3 1.4.6.4 Bechtel

................................................................................. 1.4-3 1.4.6.5 Deleted ................................................................................ 1.4-3 1.4.6.6 Westinghouse Electric

.............................................................. 1.4-3 1.4.7 CONSULTING ENGINEER - R. W. BECK AND ASSOCIATES ........... 1.4-3

1.5 REQUIREMENTS

FOR FURTHER TECHNICAL INFORMATION .......... 1.5-1 1.5.1 GENERIC ISSUES ..................................................................... 1.5-1 1.5.1.1 Unresolved Safety Issues ........................................................... 1.5-2 1.5.1.1.1 Unresolved Safety Issues Introduction

......................................... 1.5-2 1.5.1.1.2 Implementation of Specific Unresolved Safety Issues ...................... 1.5-2 1.5.1.1.3 Unresolved Safety Issues Implementation Summary ........................ 1.5-5 1.5.1.2 Generic Safety Issues ............................................................... 1.5-6 1.5.1.2.1 Generic Safety Issues Introduction ............................................. 1.5-6 1.5.1.2.2 Implementation of Specific Generic Safety Issues ........................... 1.5-6 1.5.1.2.3 Generic Safety Issues Implementation Summary

............................ 1.5-8 1.5.1.3 TMI Task Action Plans ............................................................. 1.5-8 1.

5.2 REFERENCES

.......................................................................... 1.5-8

1.6 MATERIAL

INCORPORATED BY REFERENCE

................................. 1.6-1

1.7 ACRONYMS

............................................................................... 1.7-1 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

TABLE OF CONTENTS (Continued)

Section Page LDCN-08-004 1-vi 1.8 CONFORMANCE TO NRC REGULATORY GUIDES...........................1.8-1 1.

8.1 INTRODUCTION

......................................................................1.8-1 1.8.2 NUCLEAR STEAM SUPPLY SYSTEM SCOPE OF SUPPLY EVALUATION.........................................................................1.8-1 1.8.3 BALANCE OF PLANT SCOPE OF SUPPLY EVALUATION...............1.8-87

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

LIST OF TABLES

Number Title Page 1-vii 1.2-1 Principal Regulations and Codes Followed in Plant Design.............1.2-39 1.2-2 Plant Shielding and Zone Classification....................................1.2-40 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics...................................................................1.3-3

1.3-2 Comparison of Power Conversion System Design Characteristics...................................................................1.3-9

1.3-3 Comparison of Engineered Safety Features Design Characteristics...................................................................

1.3-10 1.3-4 Comparison of Containment Design Characteristics......................1.3-12

1.3-5 Radioactive Waste Management Systems Design Characteristics...................................................................

1.3-14 1.3-6 Comparison of Structural Design Characteristics.........................1.3-15

1.3-7 Comparison of Electrical Systems Design Characteristics...............1.3-16

1.3-8 Significant Design Changes from PSAR to FSAR........................1.3-17

1.4-1 Commercial Nuclear Reactors Completed, Under Construction,

or in Design by General Electric.............................................1.4-5

1.6-1 Topical Reports.................................................................1.6-3

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

LIST OF FIGURES

Number Title LDCN-98-117 1-viii 1.2-1 Plant Plot Plan 1.2-2 General Arrangement - Ground Floor Plan, Turbine Generator Building (Sheets 1 and 2) 1.2-3 General Arrangement - Mezzanine Floor Plan, Turbine Generator Building (Sheets 1 and 2)

1.2-4 General Arrangement - Operating Floor Plan, Turbine Generator Building (Sheets 1 and 2)

1.2-5 General Arrangement - Sections 2-2, 4-4 and 5-5, Turbine Generator Building

1.2-6 General Arrangement - Sections 1-1 and 3-3, Turbine Generator Building 1.2-7 General Arrangement - El. 422 ft 3 in., El. 441 ft 0 in., and 444 ft 0 in., Reactor Building 1.2-8 General Arrangement - El. 471 ft 0 in. and El. 501 ft 0 in., Reactor Building

1.2-9 General Arrangement - El. 522 ft 0 in. and El. 548 ft 0 in., Reactor Building

1.2-10 General Arrangement - El. 572 ft 0 in. and El. 606 ft 10-1/2 in., Reactor Building 1.2-11 General Arrangement - Section 10-10, Reactor Building

1.2-12 General Arrangement 8 and 9-9, Reactor Building

1.2-13 General Arrangement - El. 437 ft 0 in., Radwaste Building

1.2-14 General Arrangement - El. 467 ft 0 in. and Partial Plans, Radwaste Building

1.2-15 General Arrangement - El. 484 ft 0 in., El. 487 ft 0 in., and Partial Plans, Radwaste Building

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

LIST OF FIGURES (Continued)

Number Title LDCN-02-000 1-ix 1.2-16 General Arrangement - El. 501 ft 0 in., 507 ft 0 in., and 525 ft 0 in., Radwaste Building 1.2-17 General Arrangement - Radwaste Building Sections (Sheets 1 and 2)

1.2-18 General Arrangement - Service Building

1.2-19 General Arrangement - Service Building Sections

1.2-20 General Arrangement - Standby Service Water Pump Houses Sections

1.2-21 General Arrangement - Circulating Water Pump House Sections

1.2-22 General Arrangement - Diesel Generator and Service Building Sections

1.2-23 General Arrangement - Makeup Water Pump House, Plans and Sections

1.2-24 General Arrangement - Makeup Water Pump House, Plans and Sections

1.2-25 General Electric Piping and Instrumentation Drawing Symbols

1.2-26 Flow Diagram Legend, Symbols and Abbreviations

1.2-27 System Acronyms

1.2-28 Equipment Acronyms (Sheets 1 and 2)

1.2-29 Logic Symbols for NSSS Functional Control Diagrams (Sheets 1 through 15)

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-005 1.1-1 Chapter 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

This Final Safety Analysis Report (FSAR) was submitted in support of an application by

Energy Northwest for a Class 103 operating license for a single unit nuclear power plant. The facility is known as the Columbia Generating Station (CGS) and was formerly known as WNP-2.

Energy Northwest was the applicant for the operating license for CGS. The plant was designed, constructed, and is being operated under the responsibility of Energy Northwest.

CGS is located within the Hanford Site of the Department of Energy (DOE), Benton County, Washington, approximately 12 miles north of the City of Richland. The site is approximately

3 miles west of the Columbia River at River Mile 352.

This plant has a boiling water reactor (BWR) nuclear steam supply system (NSSS) designed and supplied by the General Electric Company (GE). The plant utilizes a single-cycle, forced-

circulation system and is designated as a BWR/5.

The containment was designed by Burns and Roe, Inc., and consists of primary and secondary containment systems. The primary containment structure is a free-standing steel pressure vessel of a specific design by Pittsburgh Des Moines Steel Co. The vessel contains both a

drywell and a suppression chamber, which is consistent with the features of a BWR/Mark II

containment.

The secondary containment structure is composed of the reactor building, which completely

encloses primary containment.

The authorized maximum rated power level limit of the reactor is 3544 MWt. The design power level limit is 3629 MWt. The net electrical power output is approximately 1190 MWe and the gross electrical output is 1230 MWe.

Energy Northwest was granted an operating license for CGS on December 20, 1983, and the plant began commercial operation on December 13, 1984.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.2-1 1.2 GENERAL PLANT DESCRIPTION

1.2.1 PRINCIPAL

DESIGN CRITERIA

The principal design criteria are presented in two ways.

First, they are classified as either a power generation function or a safety function. Second, they are grouped according to system. Although the distinctions between power generation or safety functions are not always clear-cut and are sometimes overlapping, the functional classification facilitates safety analyses, while the grouping by system facilitates the understanding of both the system function and design.

1.2.1.1 General Design Criteria

1.2.1.1.1 Power Generation Design Criteria

a. The plant was designed so that it can be fabricated, erected, and operated to produce electric power in a safe and reliable manner. Plant design conforms to

applicable codes and regulations as stipulated in Table 1.2-1

b. The plant is designed to produce steam for direct use in a turbine-generator unit;

c. Heat removal systems are provided with sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions and abnormal operational transients;

d. Backup heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal

systems become inoperative. The capacity of such systems is adequate to

prevent fuel cladding damage;

e. The fuel cladding, in conjunction with other plant systems is designed to retain integrity throughout the range of normal operational conditions and abnormal

operational transients;

f. The fuel cladding can accommodate, without loss of integrity, the pressures generated by fission gases released from fuel material throughout the design life

of fuel;

g. Control equipment has been provided to allow the reactor to respond automatically to minor load changes, major load changes, and abnormal

operational transients;

h. Reactor power level can be manually controlled; C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.2-2 i. Control of the reactor is possible from a single location;

j. Reactor controls, including alarms, are arranged to allow the operator to rapidly assess the condition of the reactor syst em and locate system malfunctions; and

k. Interlocks or other automatic equipment are provided as backup to procedural controls to avoid conditions requiring the functioning of nuclear safety systems

or engineered safety features (ESF).

1.2.1.1.2 Safety Design Criteria

a. The plant design conforms to applicable codes and regulations;

b. The plant is designed, fabricated, erected, and will be operated in such a way that the release of radioactive materials to the environment is limited to the

limits and guideline values of applicable federal regulations pertaining to the

release of radioactive materials for normal operations and abnormal transients

and accidents;

c. The reactor core is designed so its nuclear characteristics do not contribute to a divergent power transient;

d. The reactor is designed so there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the reactor with other appropriate plant systems;

e. Gaseous, liquid, and solid waste disposal facilities are designed so the discharge and offsite shipment of radioactive effluents can be made in accordance with

applicable regulations;

f. The design provides means by which plant operators can be informed when limits on the release of radioactive material are approached;

g. Sufficient indications are provided to allow determination that the reactor is operating within the envelope of conditions considered by plant safety analysis;

h. Radiation shielding is provided and access control patterns have been established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any mode of normal plant operations;

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November1998 1.2-3 i. Those portions of the nuclear system that form part of the reactor coolant pressure boundary (RCPB) are designed to retain integrity as a radioactive material barrier following abnormal operational transients and accidents;

j. Nuclear safety systems and ESF act to ensure that no damage to the RCPB results from internal pressures caused by abnormal operational transients and

accidents;

k. Where positive, precise action is immediately required in response to abnormal operational transients and accidents, such action is automatic and requires no decision or manipulation of controls by plant operations personnel;

l. Essential safety actions can be carried out by equipment of sufficient redundance and independence such that no single failure of active components can prevent the required actions. For systems or components to which IEEE-279 (Criteria for Protection Systems for Nuclear Power Generating Stations) and/or IEEE-308 (Criteria for Class 1E Electrical systems for Nuclear Power Generating Stations) applies, single failures of both active and passive electrical components were considered in recognition of the higher anticipated failure rates of passive

electrical components relative to passive mechanical components;

m. Provisions have been made for control of active components of nuclear safety systems and ESF from the control room;

n. Nuclear safety systems and ESF are designed to permit demonstration of their functional performance requirements;

o. The design of nuclear safety systems and ESF includes allowances for natural environmental disturbances such as earthquakes, tornadoes, floods, and storms

at the site;

p. Standby electrical power sources have sufficient capacity to power all nuclear safety systems and ESF requiring electrical power;

q. Standby electrical power sources are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where offsite power sources are

not available;

r. Features of the plant that are essential to the mitigation of accident consequences are designed, fabricated, and erected to quality standards that reflect the importance of the safety action to be performed;

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-05-009 1.2-4 s. A primary containment has been provided that completely encloses the reactor system, drywell, and suppression pool. The primary containment employs the

pressure suppression concept;

t. The primary containment is designed to retain integrity as a radioactive material barrier during and following accidents that release radioactive material into the

primary containment volume;

u. It is possible to test primary containment integrity and leaktightness at periodic intervals;

v. A secondary containment has been provided that completely encloses both the primary containment and fuel storage areas. The secondary containment includes the standby gas treatment (SGT) system for controlling release of radioactive materials leaking from the primary containment in the event of an accident and also has the capability for filtering radioactive materials directly from the primary containment atmosphere during shutdown conditions;

w. The secondary containment has been designed to act as a radioactive material barrier, if required, when the primary containment is open for expected

operational purposes;

x. The primary containment and secondary containment, in conjunction with other ESF, limit radiological effects of accidents resulting in the release of radioactive material to the containment vessel to significantly less than 10 CFR 50.67 limits;

y. Provisions have been made for removing energy from within the containment vessel as necessary to maintain the integrity of the containment system following accidents that release energy to the primary containment;

z. Piping that penetrates the primary containment structure and could serve as a path for the uncontrolled release of radioactive material to the environs is automatically isolated whenever such uncontrolled radioactive material release is threatened. Such isolation shall be effected in time to limit radiological effects to less than specified acceptable limits; aa. Emergency core cooling systems (ECCS) are provided to limit fuel cladding temperature to temperatures below the onset of fragmentation in the event of a

loss-of-coolant accident (LOCA);

bb. The ECCS provide for continuity of core cooling over the complete range of postulated break sizes in the RCPB and are redundant; C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 1.2-5 cc. Operation of the ECCS is initiated automatically when required, regardless of the availability of offsite power supplies and the normal generating system of

the plant;

dd. The control room has been shielded against radiation and provided with a high efficiency filtration system so that continued occupancy under accident conditions is possible; ee. In the event that the control room becomes inaccessible, it is possible to bring the reactor from power range operation to cold shutdown conditions by utilizing the local controls and equipment that are available outside the control room on

the remote shutdown control panels;

ff. Backup reactor shutdown capability has been provided independent of normal reactivity control provisions. This backup system has the capability to shut down the reactor from any normal operating condition and subsequently to maintain the shutdown condition; and

gg. Fuel handling and storage facilities are designed to prevent inadvertent criticality and to maintain adequate shielding and cooling of spent fuel.

Provision is made for maintaining the cleanliness of spent fuel cooling and

shielding water.

1.2.1.2 System Criteria

The principal design criteria for particular systems are listed in the following subsections.

1.2.1.2.1 Nuclear System Criteria

a. The fuel cladding is designed to retain integrity as a radioactive material barrier throughout the design power range. The fuel cladding is designed to accommodate, without loss of integrity, the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel;

b. The fuel cladding, in conjunction with other plant systems, is designed to retain integrity throughout any abnormal operational transient;

c. Those portions of the nuclear system that form part of the RCPB are designed to retain integrity as a radioactive material barrier following abnormal operational transients and accidents;

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 1.2-6 d. Heat removal systems are provided in sufficient capacity and operational adequacy to remove heat generated in the reactor for the full range of normal operational conditions from plant shutdown to design power and for any abnormal operational transient. The capacity of such systems is adequate to

prevent fuel cladding damage;

e. Heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems

become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage. The reactor is capable of being automatically shut down in sufficient time to permit decay heat removal systems to become effective following loss of operation of normal heat removal systems;

f. The reactor core and reactivity control system is designed so that control rod action is capable of bringing the core subcritical and maintaining it so, even with the rod of highest reactivity worth fully withdrawn and unavailable for

insertion;

g. The reactor core is designed so that its nuclear characteristics do not contribute to a divergent power transient; and

h. The nuclear system is designed so there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the nuclear system

with other appropriate plant systems.

1.2.1.2.2 Power Conversion System Criteria

Components of the power conversion system have been designed to perform the following

basic objectives.

a. Produce electrical power from the steam exiting from the reactor, condense the steam into water, and return the water to the reactor as heated feedwater, with a major portion of its gaseous and particulate impurities removed; and

b. Ensure that any fission products or radioactivity associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal

procedures.

1.2.1.2.3 Electrical Power Systems Criteria

Sufficient offsite and onsite standby sources of electrical power are provided to attain prompt shutdown and continued maintenance of the plant in a safe condition under all credible C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 1.2-7 circumstances. The power sources are adequate to accomplish all required engineered safety feature functions under postulated design basis accident conditions.

1.2.1.2.4 Radwaste System Criteria

a. The gaseous and liquid radwaste systems are designed to limit the release of radioactive effluents from the plant during normal operation within those limits

specified in 10 CFR 20 and 10 CFR 50, Appendix I;

b. The solid radwaste disposal system is designed so that during normal operation offsite shipments will be in accordance with applicable regulations, including 10 CFR 20, 10 CFR 71, and 49 CFR 171 through 10 CFR 179, as appropriate;

and

c. The design of the systems provide means by which plant operations personnel are alerted whenever operational limits on the release of radioactive material are

approached.

1.2.1.2.5 Auxiliary Systems Criteria

a. Fuel handling and storage facilities are designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel. Provision is made for maintaining the cleanliness of spent fuel cooling and shielding water;

b. Other auxiliary systems, such as standby service water (SW), high pressure core spray (HPCS) SW, fire protection (FP), heating and ventilating, communications, and lighting systems, are designed to function during normal, abnormal, and/or accident conditions; and

c. Auxiliary systems that are not required to effect safe shutdown of the reactor or maintain it in a safe shutdown condition are designed such that failure of these systems shall not prevent the essential auxiliary systems from performing their

design functions.

1.2.1.2.6 Shielding and Access Control Criteria

a. Radiation shielding is provided and access control patterns are established to allow a properly trained operating staff to control radiation doses within the limits of published regulations in any normal mode of plant operation; and

b. The control room is shielded against radiation and has a high efficiency filtration system, so that occupancy is possible under accident conditions and C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-05-009 1.2-8 TEDE doses are less than those set by Criterion 19 of 10 CFR Part 50, Appendix A and 10 CFR 50.67.

1.2.1.2.7 Nuclear Safety Systems and ESF Criteria

Principal design criteria for nuclear safety systems and ESF correspond to criteria j through q, aa through cc, and ee through ff in Section 1.2.1.1.2.

1.2.1.2.8 Process Control Systems Criteria The principal design criteria for the process control systems are listed for the nuclear system, the power conversion system, and the electrical power system:

a. Nuclear System Process Control Criteria

1. Control equipment is provided to allow the reactor to respond automatically to load changes within design limits.

2. It is possible to manually control the reactor power level.

3. Control of the reactor is possible from a central location.

4. Nuclear systems process controls and alarms are arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

5. Interlocks or other automatic equipment are provided as a backup to procedural controls to avoid conditions requiring the actuation of nuclear

safety systems or ESF.

b. Power Conversion System Process Control Criteria

1. Control equipment is provided to control the reactor pressure throughout its operating range.

2. The turbine is able to respond automatically to minor changes in load.

3. Control equipment in the feedwater system maintains the water level in the reactor vessel at the optimum level required by steam separators.

4. Control of the power conversion equipment is possible from a central location.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 1.2-9 5. Interlocks or other automatic equipment are provided in addition to procedural controls to avoid conditions requiring the actuation of ESF.

c. Electrical Power System Process Control Criteria

1. The redundant portions of the Class 1E power systems are designed with either division of the system being adequate to safely shut down the unit.

2. Protective relaying is used to detect and isolate faulted equipment from the system with a minimum of disturbance in the event of equipment

failure.

3. Primary and secondary undervoltage relays are located on the 4.16-kV Class 1E equipment buses to isolate these buses from the normal auxiliary power system in the event of Class 1E bus under voltage and to

initiate starting of the standby power system diesel generators.

4. Standby power diesel generators start is initiated by control relays. The generators are also loaded by a sequenced control system to meet the

existing emergency condition.

5. All electrically operated breakers can be operated from the main control room.

6. Metering for essential generators, transformers, and circuits is monitored in the main control room.

1.2.1.3 Plant Design Criteria

The plant design criteria are based on general design criteria given in Appendix A of 10 CFR Part 50. Conformance to these criteria is discussed in Section 3.1. The classification of structures, components, and systems is discussed in Section 3.2.

The principal regulations are codes that are used extensively in plant design are highlighted in

Table 1.2-1. Note that the codes listed may not be applicable in their entirety. The many codes and regulations applicable to individual systems or structures are discussed throughout

the FSAR.

The plant shielding and radiation zone classification can be found in Table 1.2-2. Chapter 12 provides further details.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December2005 LDCN-04-052 1.2-10 1.2.2 PLANT DESCRIPTION

1.2.2.1 Site Characteristics

1.2.2.1.1 Site Location and Size

Columbia Generating Station (CGS) is located in the southeast area of the Department of Energy (DOE) Hanford Reservation in Benton County, Washington. The site is approximately 3 miles west of the Columbia River at River Mile 352, approximately 12 miles north of the City of Richland, 18 miles northwest of Pasco, and 21 miles northwest of Kennewick. The site is approximately square shaped with a corridor extending to the makeup water pump house

located on the Columbia River as shown in Figure 1.2-1. The CGS site encompasses an area of approximately 1089 acres.

1.2.2.1.2 Description of Site Environs

1.2.2.1.2.1 Site Land. See Section 2.1 for site land description.

1.2.2.1.2.2 Population. See Section 2.1 for population description.

1.2.2.1.2.3 Land Use. Natural physical characteristics of the site which make it well-suited for operation of the plant include: favorable geographical, geological, and seismological characteristics; adequate water supply; ideal climatological characteristics; and remoteness from population centers or areas of special ecological concern. The site area had served as a nuclear industrial center since 1943 when it was selected by the federal government as the location for construction of one of the worlds first nuclear production reactors. Since 1943, nine plutonium production reactors and a number of test reactors have been constructed and

operated at the Hanford Site.

1.2.2.1.2.4 Meteorology. The climate around CGS is basically continental with a wide range of annual temperatures. See Section 2.3 for additional information.

1.2.2.1.2.5 Hydrology. The Columbia River is the major surface water resource of the region. The river also forms a potential discharge boundary for the aquifer. The surface soils at Hanford are sufficiently permeable to take in water from precipitation and industrial discharges. See Section 2.4 for additional information.

1.2.2.1.2.6 Geology. The Hanford site lies in the east central part of the Pasco Basin, a structural and topographic depression in the Columbia Plateau. The region is underlain by three major geologic units: (a) Tertiary basaltic lavas and intercalated sediments of the Columbia River Group at the base, (b) Plio-Pleistocene sediments of the Ringold Formation, and (c) the Pasco (glaciofluvial) gravels and associated sediments of late Pleistocene age at the surface. See Section 2.5 for additional information.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-052 1.2-11 1.2.2.1.2.7 Seismology. The CGS site is situated in an area characterized by low seismicity and widely scattered epicenters. See Section 2.5 for additional information.

1.2.2.1.3 Design Basis Depending on Site Environs

a. Offgas System

An offgas (OG) system consisting of hold-up piping, charcoal adsorbers, and an elevated release is provided for the controlled release of gaseous effluent to the atmosphere. Gaseous releases will be as low as reasonably achievable (ALARA) in accordance with 10 CFR Part 50, Appendix I, and less than

10 CFR Part 20 limits;

b. Liquid Waste Effluents

Liquid waste will be processed and recycled, and releases of excess inventory will be such that concentrations at the point of discharge will be as low as

reasonably achievable in accordance with 10 CFR Part 50, Appendix I, and less

than 10 CFR Part 20 limits;

c. Wind Loading and Seismic Design The structures and components whose failure might cause a design basis accident or result in an uncontrolled release of radioactive fission products will be designed to resist wind loads of tornado velocity and earthquake ground motions which are significantly higher than those expected to occur at the site during the service life of the plant; and

d. Flooding

The maximum assumed flood elevation for design purposes is the sum total of the elevations of water due to the following effects:

1. Breach of any of the upstream dams due to seismic forces, 2. High flow in the Columbia River, and 3. Wind and wave action.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-052 1.2-12 1.2.2.2 General Arrangement of Structures and Equipment

The principal structures located on the plant site are the following:

a. Reactor building - the building that houses the major portion of the nuclear steam supply system (NSSS), the drywell, suppression pool, primary containment, new and spent fuel pools, refueling equipment, and ECCS;

b. Radwaste and control building - the building that houses the liquid and solids radwaste systems, components of the OG system, and the main control room;

c. Turbine building - the building that houses the power conversion equipment;

d. Diesel generator building - the building that houses the standby diesel generators, diesel fuel oil (DO) storage tanks, and associated controls and instrumentation;

e. Circulating water pump house (Wind River Building) - a structure housing the main circulating water (CW) pumps, plant service water (TSW) pumps, and FP pumps;

f. Standby service water pump houses - structures that house the redundant standby SW pumps and the HPCS SW pump;

g. Spray ponds - cooling ponds provided as the ultimate heat sink (UHS);

h. Makeup water pump house - a structure that houses the cooling tower makeup (TMU) water pumps;

i. General service building (Yakima Building) - a structure that houses the potable water (PWC) storage tank, demineralized water (DW) storage tank, offices for plant administration, lunch room, and machine shop;

j. Transformer yard;

k. Condensate storage tanks (CSTs);

l. Cooling towers; and

m. Plant Engineering Center (Deschutes Building).

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-052 1.2-13 The arrangement of these structures on the plant site is shown in Figure 1.2-1. The arrangement of the equipment inside the main buildings is shown in Figures 1.2-2 through 1.2-24.

1.2.2.3 Symbols Used on Engineering Drawings

Figure 1.2-25 defines General Electrics (GE) piping and instrumentation symbols, and Figure 1.2-26 through 1.2-28 shows Burns and Roe piping and instrumentation symbols.

Figure 1.2-29 defines the logic symbols used on NSSS functional control diagrams.

1.2.2.4 Nuclear System

The nuclear system includes a direct-cycle, forced-circulation, GE boiling water reactor (BWR) that produces steam for direct use in the steam turbine. A heat balance showing the

major parameters of the nuclear system for the rated power conditions is shown in

Figure 10.1-1.

1.2.2.4.1 Reactor Core and Control Rods

Fuel for the reactor core consists of slightly enriched uranium dioxide pellets sealed in Zircaloy-2 tubes. These tubes (or fuel rods) are assembled into individual fuel assemblies.

Gross control of the core is achieved by movable, bottom-entry control rods. The control rods are cruciform in shape and are dispersed throughout the lattice of fuel assemblies. The control rods are positioned by individual control rod drives (CRDs).

Each fuel assembly has several fuel rods with gadolinia (Gd 2 O 3) mixed in solid solution with UO 2. The Gd 2 O 3 is a burnable poison which diminishes the reactivity of the fresh fuel. It is depleted as the fuel reaches the end of its first cycle.

A conservative limit of plastic strain is the design criterion used for fuel rod cladding failure.

The peak linear heat generation for steady-state operation is well below the fuel damage limit even late in life. Experience has shown that the control rods are not susceptible to distortion and have an average life expectancy many times the residence time of the fuel loading.

1.2.2.4.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structures; the steam separators and dryers; the jet pumps; the control rod guide tubes; the distribution lines for reactor feedwater (RFW), HPCS, low-pressure core spray (LPCS), and standby liquid control (SLC); the in-core instrumentation; and other components. The main connections to the vessel include main steam (MS) lines, reactor recirculation (RRC) lines, RFW lines, CRD and in-core nuclear instrument housings, HPCS and LPCS lines, residual heat removal (RHR) lines, SLC line, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-04-052 1.2-14 core differential pressure line, jet pump pressure-sensing lines, and water level

instrumentation.

The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure in the steam space above the separators

is 1035 psia. The vessel is fabricated of low-alloy steel and is clad internally with stainless steel (except for the top head, and certain nozzles and nozzle weld zones which are unclad).

The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators and dryers located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the MS lines. Each MS line is provided with two MS isolation valves (MSIVs) in series, one on each side of the primary containment barrier.

1.2.2.4.3 Reactor Recirculation System

The RRC system pumps reactor coolant through the core. This is accomplished by two recirculation loops external to the reactor vessel but inside the primary containment. Each external loop contains a mechanical pump, two motor-operated maintenance valves, and one flow control valve which is mechanically blocked full open. The two motor-operated valves are used as pump suction and pump discharge shutoff valves. The flow control valves are no longer used to control reactor power level and therefore are kept in a mechanically blocked full

open position.

The internal portion of the loop consists of the jet pumps, which contain no moving parts. The jet pumps provide a continuous internal circulation path for the major portion of the core coolant flow. The jet pumps are located in the annular region between the core shroud and the vessels inner wall. Any recirculation line break would still allow core flooding to

approximately two-thirds of the core height, the level of the inlet of the jet pumps.

1.2.2.4.4 Residual Heat Removal System

The RHR system is a system of pumps, heat exchangers, and piping that fulfills the following

functions:

a. Removes decay and sensible heat during and after plant shutdown;

b. Injects water into the reactor vessel, following a LOCA, rapidly enough to reflood the core and maintain fuel cladding below the fragmentation temperature

independent of other core cooling systems. This is further discussed in

Section 1.2.2.5.8; C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-05-009 1.2-15 c. Removes heat from the primary containment, following a LOCA, to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the suppression pool water (containment cooling) and by spraying

the drywell and suppression pool air spaces (containment spray) with

suppression pool water; and

d. Removes some of the airborne radioactivity from the primary containment atmosphere following a LOCA by spraying the drywell.

1.2.2.4.5 Reactor Water Cleanup System

The reactor water cleanup (RWCU) system recirculates a portion of reactor coolant through a filter-demineralizer to remove particulate and dissolved impurities from the reactor system under controlled conditions. It also removes excess coolant from the reactor system under

controlled conditions.

1.2.2.4.6 Nuclear Leak Detection System

The nuclear leak detection (LD) system consists of temperature, pressure, flow, and fission-product sensors with associated instrumentation and alarms. This system detects and

annunciates leakage in the following systems:

a. Main steam system,

b. Reactor water cleanup system,

c. Residual heat removal system,

d. Reactor core isolation cooling (RCIC) system,

e. Reactor feedwater system,

f. High-pressure core spray system,

g. Low-pressure core spray system,

h. Reactor recirculation system, and

i. Reactor pressure vessel (RPV) flange.

Small leaks generally are detected by temperature and pressure changes, fill-up rate of drain sumps, and fission-product concentration inside the primary containment. Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines.

1.2.2.5 Nuclear Safety Systems and Engineered Safety Features

1.2.2.5.1 Reactor Protection System

The reactor protection system (RPS) initiates a rapid, automatic shutdown (scram) of the reactor, if required, to prevent fuel cladding damage or nuclear system process barrier damage following abnormal operational transients. The RPS overrides all operator actions and process C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December2005 LDCN-04-052 1.2-16 controls and is based on a fail-safe design philosophy that allows appropriate protective action even if a single component failure occurs.

1.2.2.5.2 Neutron Monitoring System

Although not all portions of the neutron monitoring system qualify as a nuclear safety system, those that provide high neutron flux signals to the RPS do. The intermediate range monitors (IRMs) and average power range monitors (APRMs), which monitor neutron flux via in-core detectors, signal the RPS to scram in time to prevent excessive fuel cladding damage as a result of overpower transients. The APRM modules also provide inputs to the thermal power monitors (TPMs) which approximate fuel thermal conditions and also provide scram signals to

the RPS.

1.2.2.5.3 Control Rod Drive System

When a scram is initiated by the RPS, the CRD system inserts the negative reactivity necessary to shut down the reactor. Each control rod is controlled individually by a hydraulic control unit. When a scram signal is received, high-pressure water stored in an accumulator in the hydraulic control unit forces its control rod into the core.

1.2.2.5.4 Control Rod Drive Housing Supports

Control rod drive housing supports are located underneath the reactor vessel near the control

rod housings. The supports limit the travel of a control rod in the event that a control rod housing is ruptured. The supports prevent a nuclear excursion as a result of a housing failure

and thus protect the fuel barrier.

1.2.2.5.5 Control Rod Velocity Limiter

A control rod velocity limiter is attached to each control rod to limit the velocity at which a control rod can fall out of the core should it become detached from its CRD. This action limits the rate of reactivity insertion resulting from a rod drop accident. The limiters contain no moving parts.

1.2.2.5.6 Pressure Relief System (Nuclear System)

A pressure relief system consisting of safety/relief valves (SRVs) mounted on the MS lines is provided to prevent excessive pressure inside the nuclear system following either abnormal

operational transients or accidents.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-14-030 1.2-17 1.2.2.5.7 Reactor Core Isolation Cooling System The RCIC system provides makeup water to the reactor vessel when the vessel is isolated.

The RCIC system uses a steam-driven turbine-pump unit and operates automatically in time and with sufficient coolant flow to maintain adequate water level in the reactor vessel. RCIC is not an engineered safety feature. It is included here, however, because of its similar functions.

1.2.2.5.8 Emergency Core Cooling System Four ECCS are provided to maintain fuel cladding below fragmentation temperature in the

event of a breach in the RCPB that results in a loss of reactor coolant. The systems are

a. High-pressure core spray system,

b. Automatic depressurization system (ADS),

c. Low-pressure core spray system, and

d. Low-pressure coolant injection (LPCI), an operating mode of the RHR system.

1.2.2.5.8.1 High-Pressure Core Spray System. The HPCS system provides and maintains an adequate coolant inventory inside the reactor vessel to maintain fuel cladding temperatures below the fragmentation temperature in the event of breaks in the RCPB. The system is initiated by either high pressure in the drywell or low water level in the vessel. It operates independently of all other systems over the entire range of pressure differences from greater than normal operating pressure to zero. The HPCS cooling decreases vessel pressure to enable the low pressure cooling systems to function. The HPCS system is powered by its own diesel generator if auxiliary power is not available, and the system may also be used as a backup for

the RCIC system.

1.2.2.5.8.2 Automatic Depressurization System. The ADS rapidly reduces reactor vessel pressure during a LOCA situation in which the HPCS system fails to maintain the reactor vessel water level. The depressurization provided by the system enables the low pressure ECCS to deliver cooling water to the reactor vessel. The ADS uses some of the relief valves that are part of the nuclear system pressure relief system. The automatic relief valves are arranged to open when conditions indicate that the HPCS system is not delivering sufficient cooling water to the reactor vessel to maintain the water level above a preselected value. The ADS will not be activated unless either the LPCS or LPCI pumps are operating. This is to ensure that adequate coolant will be available to maintain reactor water level after the depressurization.

1.2.2.5.8.3 Low-Pressure Core Spray System. The LPCS system consists of one independent pump and the valves and piping to deliver cooling water to a spray sparger over the core. The

system is actuated by conditions indicating that a breach exists in the RCPB but water is delivered to the core only after reactor vessel pressure is reduced. This system provides the capability to cool the fuel by spraying water into each fuel channel. The LPCS loop C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-05-009 1.2-18 functioning in conjunction with either the ADS or HPCS can maintain the fuel cladding below the prescribed temperature following a LOCA.

1.2.2.5.8.4 Low-Pressure Coolant Injection. The LPCI is an operating mode of the RHR system, but is discussed here because the LPCI mode acts as an engineered safety feature in

conjunction with other ECCS. The LPCI uses the pump loops of the RHR to inject cooling water directly into the pressure vessel. The LPCI is actuated by conditions indicating a breach in the RCPB, but water is delivered to the core only after reactor vessel pressure is reduced.

The LPCI operation provides the capability of core reflooding, following a LOCA, in time to maintain the fuel cladding below the prescribed temperature limit.

1.2.2.5.9 Primary Containment

1.2.2.5.9.1 Functional Design. The primary containment is part of the overall containment system which provides the capability to reliably limit the release of radioactive materials to the environs subsequent to the occurrence of the postulated LOCA so that offsite doses will be below the limits stated in 10 CFR Part 50.67. Its design employs an over-and-under, steel pressure vessel which houses the reactor vessel, the RRC loops, and other branch connections

of the reactor primary system. The pressure suppression system consists of a drywell, a pressure suppression chamber which stores a large volume of water, a connecting submerged vent system between the drywell and water pool, isolation valves, containment cooling system, and other service equipment. In the event of a RCPB piping failure within the drywell, reactor water and steam would be released into the drywell air space. The resulting increase of drywell pressure would then force a mixture of air, steam, and water through the vents into the

pool of water which is stored in the suppression pool, resulting in a rapid pressure reduction in the drywell. Air which is transferred to the suppression chamber, pressurizes the suppression

chamber, and is subsequently vented back to the drywell.

1.2.2.5.9.2 Drywell Cooling System. The drywell cooling system is based on recirculating cooling water through the drywell air-handling units to maintain the required ambient

temperature. Air is distributed through ductwork and/or up through the annular space between the reactor vessel insulation and the sacrificial shield wall. Air is distributed to areas requiring cooling, such as the RRC motors, the CRD area, and the bellows area. Return air is ducted back to the operating units. The arrangement simplifies the design, operation, and air

distribution balance of the system.

Reactor building closed cooling water (RCC) is supplied to the air handling units to dissipate absorbed heat only under normal and loss of power conditions.

The drywell cooling system is not required for safe shutdown, but it is designed with redundant

equipment and powered from essential buses to ensure continuous operation to satisfy the

power-generation design objective.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-05-009 1.2-19 The drywell cooling system is designed to operate during offsite power loss. Control switches

for operating the equipment are located in the main control room.

1.2.2.5.9.3 Suppression Pool Cooling. The containment cooling subsystem of the RHR system is placed in operation to limit the temperature of the water in the suppression pool following a design basis LOCA, to control the pool temperature during normal operation of the SRVs and the RCIC system, and to reduce the pool temperature following an isolation transient. In the containment cooling mode of operation, the RHR main system pumps take

suction from the suppression pool and pump the water through the RHR heat exchangers where cooling takes place by transferring heat to SW. The fluid is then discharged back to the suppression pool or the RPV.

1.2.2.5.9.4 Containment Spray. The redundant containment spray cooling subsystems of the RHR system provide containment cooling for postaccident conditions. Water pumped through the RHR heat exchangers can be diverted to spray headers in the drywell and above the

suppression pool. The spray removes energy from the drywell atmosphere by condensing the water vapor. The drywell spray also removes particulate fission product from the drywell atmosphere. Approximately 5% of this flow can be directed to the suppression chamber to cool the gas above the water surface.

1.2.2.5.9.5 Containment Atmosphere Control. In the event of a LOCA, hydrogen and oxygen will be generated in the reactor. Containment atmosphere control is provided by inerted containment, containment atmosphere mixing, and hydrogen and oxygen monitoring in

a post-LOCA event.

1.2.2.5.10 Primary Containment and Reactor Vessel Isolation System

The primary containment and reactor vessel isolation system includes sensors, trip channels, control switches and remotely activated valve closing mechanisms associated with the valves, which, when closed, effect isolation of the primary containment or reactor vessel or both.

The purpose of the system is to provide timely protection against the onset and consequences of accidents involving the gross release of radioactive materials from the fuel and the nuclear

system process barrier. The primary containment and reactor vessel isolation control system

initiates automatic isolation of the RCPB and the primary containment vessel whenever monitored variables exceed preselected operation limits.

All pipelines that both penetrate the primary containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-13-008 1.2-20 1.2.2.5.11 Main Steam Line Isolation Valves

Although all pipelines that both penetrate the containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities, the main steam lines, because of their large size and large mass flow rates, are given special isolation consideration.

Automatic MSIVs are provided in each MS line. Each is powered by both air pressure and spring force. These valves fulfill the following objectives:

a. Prevent excessive damage to the fuel barrier by limiting the loss of reactor coolant from the reactor vessel resulting from either a major leak from the steam piping outside the primary containment or from a malfunction of the pressure control system resulting in excessive steam flow from the reactor

vessel, b. Limit the release of radioactive materials (i.e., iodine spiking) by isolating the RCPB in case of a rapid depressurization of RPV and resulting release of radioactive materials from the fuel to the reactor cooling water and steam, and

c. Limit the release of radioactive materials by closing the primary containment barrier in case of a major leak from the nuclear system inside the primary

containment.

1.2.2.5.12 Main Steam Line Flow Restrictors

A venturi-type flow restrictor is installed in each MS line. These devices limit the loss-of-coolant from the reactor vessel before the MSIVs are closed in case of an MS line

break outside the primary containment.

1.2.2.5.13 Main Steam Line Radiation Monitoring System

The main steam line radiation monitoring system consists of four gamma radiation monitors located externally to the main steam lines just outside the containment. The monitors are designed to detect a gross release of fission products from the fuel. On detection of high radiation, the trip signals generated by the monitors are used to initiate a closure to the reactor water sample valves, mechanical vacuum pump trip, the mechanical vacuum pump lines isolation, and alarms.

1.2.2.5.14 Standby Service Water and High-Pressure Core Spray Service Water Systems

The SW system consists of two completely redundant systems. Each system consists of a pump and piping supplying the associated RHR system heat exchanger, standby diesel generator, essential heating, ventilating, and air conditioning (HVAC) coolers, RHR pump seal coolers, SW motor bearing coolers, and sample coolers with safety grade cooling water from C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-05-009 1.2-21 the UHS spray ponds. The Division I SW system also provides cooling water to the LPCS

motor bearing cooler.

Cooling water is supplied during a postulated LOCA to the RHR heat exchangers to remove heat when the containment cooling mode of the RHR system is placed in operation. During normal operation, SW is also supplied to the RHR heat exchangers for the shutdown function

of the RHR system.

The SW is available to the shell side of the fuel pool cooling and clean up (FPC) system heat exchangers in the event that the normal cooling water supply from the RCC system becomes unavailable.

The HPCS SW system shares spray pond A with the SW system. The pump supplies cooling water to the HPCS diesel generator and the essential HVAC coolers for the HPCS diesel

generator and HPCS pump areas.

Cooling water is supplied to all diesel generator cooling systems whenever the diesel

generators are started.

1.2.2.5.15 Reactor Building - Secondary Containment

The reactor building completely surrounds the primary containment. The building provides secondary containment when the primary containment is closed and in service, and serves as the primary barrier during operations with the potential to drain the reactor vessel (OPDRV). The reactor building also houses refueling and reactor servicing equipment, new and spent fuel storage facilities, and other reactor safety and auxiliary systems. Secondary containment is not required during movement of irradiated fuel assemblies or core alterations.

The design of the reactor building includes provisions for seismic load resistance and low infiltration and exfiltration rates. The building consists of poured-in-place, reinforced-concrete exterior walls up to the refueling floor. Above this level, the building structure is steel frame with insulated metal siding with sealed joints. Access to the building is through interlocked

double doors.

1.2.2.5.16 Reactor Building Ventilation Exhaust Radiation Monitoring System The reactor building ventilation exhaust radiation monitoring system consists of a number of radiation monitors arranged to monitor the activity level of the ventilation exhaust from the reactor building and primary containment. Upon detection of high radiation, the reactor

building is automatically isolated and the SGT system is started.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-04-050 1.2-22 1.2.2.5.17 Standby Gas Treatment System

The SGT system consists of two identical filter trains. Each filter train consists of a filter unit, two fans, ductwork, and associated valves.

Either filter train may be considered as an installed spare with the other train capable of

passing the required amount of air. Either train alone is capable of exchanging the total

reactor building volume once in a 24 hr period.

Each filter unit contains electric heaters, a prefilte r, high-efficiency particulate filters (water and fire resistant), an iodine filter (high ignition temperature), and instrumentation to measure temperature and flow.

The system maintains a slightly negative internal building pressure and can process all gaseous

effluent prior to its discharge from the reactor building.

All equipment is connected to the essential buses and is started either automatically or

manually from the main control room.

1.2.2.5.18 Standby Alternating Current Power Supply System

The standby ac power supply system consists of two diesel generator sets, switchgear, and

associated distribution system equipment and auxiliaries.

These diesel generator sets are associated with redundant (Divisions 1 and 2) separation divisions; each diesel generator set serves a particular division. The capacity of each diesel generator set is sufficient to attain shutdown under both normal and LOCA conditions, in the event that both the offsite and the normal auxiliary power sources are unavailable to supply plant loads. Since load distribution is such that redundant auxiliary systems are separated by division, safe shutdown can be achieved with only one of the two diesel generators operating.

The standby ac power supply system diesel generators and associated equipment are designed to Class 1E standards and are located within Seismic Category I structures. Equipment of each division is separated so that failure of any component of one division will not jeopardize proper

functioning of the other division.

Although it is not a part of the standby ac power supply system, another independent diesel generator unit supplies ac power exclusively to the HPCS system (see Section 1.2.2.5.8.1) in the event that both the offsite and the normal auxiliary power sources are unavailable to supply

plant loads.

The HPCS diesel generator may also be cross connected to either Division 1 or to Division 2 as described in Section 8.3.1.1.7.2.1.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-05-009 1.2-23 1.2.2.5.19 Direct Current Power Supply System

The dc power supply system consists of station batteries, battery chargers, distribution

equipment, and related auxiliaries.

The dc system furnishes power at three voltage levels: 250 V, 125 V, and +24 V. The 250-V

and 125-V subsystems supply power to both Class 1E and non-Class 1E loads; the 24-V subsystem supplies power for the startup range and power range neutron monitoring systems.

The primary power sources for the system are the dc output station battery chargers. Station batteries associated with each charger operate in a float-charge configuration to ensure maintaining the batteries in a fully charged condition. In the event of loss of charger dc output, the station batteries furnish a secondary source of dc supply.

The 125-V and +24-V dc power supply subsystems are each divided into electrically and

physically independent divisions. Each battery, together with its independent battery charger, is associated with one of the segregated divisions. The batteries and their associated chargers

are located in separate rooms.

The ampere-hour capacity of each battery is capable of supplying all essential loads for a minimum of 2 hr in the event that dc output from the battery chargers is lost.

1.2.2.5.20 Standby Liquid Control System

Although not intended to provide prompt reactor shutdown, as the control rods are, the Standby Liquid Control (SLC) system provides a redundant, independent, and alternate method to bring the nuclear fission reaction to subcriticality and to maintain a subcritical condition as the reactor cools. The system makes possible an orderly and safe shutdown in the event that not enough control rods can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold, clean shutdown condition.

The SLC system is also used to maintain the suppression pool pH greater than 7.0 following a LOCA to minimize re-evolving gaseous iodine fission products to the containment atmosphere.

1.2.2.5.21 Safe Shutdown from Outside the Main Control Room

In the event that the control room becomes inaccessible, the reactor can be brought from power range operation to cold shutdown conditions by the use of local controls and equipment that are

available outside the control room.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-15-033 1.2-24 1.2.2.5.22 Main Steam Line Isolation Valve Leakage Control System (Deactivated)

The main steam line isolation valve leakage control (MSLC) system was designed to minimize the fission products which could bypass the SGT system after a LOCA. The MSLC system is not credited for accident mitigation and is no longer needed; MSLC is administratively

de-activated. Connections between MSLC and other systems are physically isolated, MSLC components are de-energized, closed, or otherwise taken out of service.

1.2.2.5.23 Fuel Pool Cooling and Cleanup System The FPC system provides for the removal of decay heat from stored spent fuel and maintains specified water temperature, purity, clarity, and level. This prevents boiling of the pool water and controls the buildup of excessive radioactive materials in the cooling water, thereby minimizing potential radiation exposure to plant personnel. The cooling portion of the system

is designed to Seismic Category I requirements and may be isolated from the Seismic Category II cleanup portion of the system by automatic Seismic Category I isolation valves which actuate on low-fuel pool water level. Normally the RCC system furnishes non-safety grade cooling water to the FPC system. If required, safety grade cooling and makeup water is available to the FPC system from the SW system.

1.2.2.5.24 Hardened Containment Vent (HCV) System The HCV system is designed to meet the requirements of NRC Order EA-13-109. The primary design objective of the HCV system is to provide sufficient venting capacity from the wetwell to prevent an overpressure failure of the containment by maintaining containment pressure below the primary containment design pressure and the primary containment pressure limit (PCPL). The HCV system is designed to operate during severe accident conditions which include the elevated temperatures, pressures, radiation levels, and concentrations of combustible gases such as hydrogen and carbon monoxide associated with accidents involving extensive core damage, including accidents involving a breach of the reactor vessel by molten core debris.

The HCV system is designed as a stand-alone system that can function without the support of other plant systems. Other than the two containment isolation valves, the remaining components such as the dedicated nitrogen and battery systems and control room and remote operating stations or instrumentation are not discussed or assumed in mitigating any Chapter 15 accidents.

1.2.2.6 Power Conversion System 1.2.2.6.1 Turbine Generator The turbine is an 1800 rpm, tandem-compound (one double-flow high-pressure turbine and three double-flow low-pressure turbines), reheat unit with an electrohydraulic governor for C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-25 normal operation. The turbine generator is provided with an emergency trip system for turbine overspeed. The rating of the turbine generator is 1,173,046 kW.

The generator is a direct-driven, three-phase, 60 Hz, 25,000 V, 1800 rpm, hydrogen inner-cooled, synchronous generator rated at 1,230 MVA at 0.975 power factor, 0.58 short circuit ratio at a maximum hydrogen pressure of 78 psig.

1.2.2.6.2 Main Steam System The MS system consists of four 26-in. diameter lines (which expand to 30-in. diameter lines inside the turbine building) extending from the outermost MSIVs to the main turbine stop valves. The use of four main steam lines permits testing of the turbine stop valves and MSIVs during station operation with only a minimum of load reduction. The design pressure and temperature of the MS system from the outermost MSIV to the turbine stop valve is 1250 psig at 575°F. Other features include drains and parts of the turbine bypass system.

1.2.2.6.3 Main Condenser The main condenser is a triple-pressure, single-pass, deaerating-type condenser with a divided water box. The condenser includes provisions for accepting up to 25% of the MS flow at design conditions from the turbine bypass system and serves as a heat sink for several other flows, such as exhaust steam from the RFW pump turbines, cascading heater drains, feedwater heater shell operating vents, and condensate pump suction vents.

1.2.2.6.4 Main Condenser Evacuation System The main condenser evacuation system is designed to remove noncondensable gases from the condenser, including air inleakage and dissociation products originating in the reactor, and to continuously exhaust them to the gaseous radwaste system during operation. The system consists of two 100%-capacity, twin-element first stage and single-element second stage steam

jet air ejector units complete with intercondensers for normal plant operation and a mechanical

vacuum pump for use during startup. Discharge from the vacuum pumps during startup is routed to the elevated release point.

1.2.2.6.5 Turbine Gland Seal System

The turbine gland seal system is designed to provide a means of preventing air leakage into or radioactive steam leakage out of the turbine. The system consists of two 100% steam evaporators, steam seal pressure regulators, steam seal header, gland seal steam condenser and

blowers, and the associated piping, valves, and instrumentation.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-26 1.2.2.6.6 Steam Bypass System and Pressure Control System A turbine bypass system is provided which passes steam directly to the main condenser under the control of the pressure regulator. Steam is bypassed to the condenser whenever the reactor steaming rate exceeds the load permitted to pass to the turbine generator. The capacity of the turbine bypass system is 25% of the turbine design steam flow. The Digital Electro-Hydraulic (DEH) control system provides main turbine control (governor) valve and bypass valve position demands so as to maintain a nearly constant reactor pressure during normal plant

operation.

1.2.2.6.7 Circulating Water System

The CW system provides the condenser with a continuous supply of cooling water. It is a closed system utilizing forced draft cooling towers. Makeup water to the system is provided from TMU pumps located in an intake structure on the Columbia River. The makeup water replaces the water lost by evaporation, drift, and blowdown.

1.2.2.6.8 Condensate and Feedwater System

The condensate and feedwater system pumps condensate from the condenser hotwell to the RPV. Condensate is pumped by three main condensate (COND) pumps through the gland seal steam condenser, the steam jet air ejector condensers, and the offgas condenser. After leaving

the offgas condenser, the condensate is pumped through a full-flow condensate filter-demineralizer system. The filter-demineralizer effluent is then pumped by three condensate booster pumps through the five low-pressure heaters. The last low-pressure heater discharges to the suction of the RFW pumps. The discharge from the two turbine-driven RFW pumps passes through the sixth stage of feedwater heating and then flows to the RPV.

Feedwater flow is controlled by varying the speed of the steam-driven turbine.

1.2.2.6.9 Condensate Filter-Demineralizer System

The full-flow condensate filter-demineralizer system with instrumentation and semiautomatic controls is designed to ensure a constant supply of high-quality water to the reactor.

1.2.2.7 Electrical Systems, Instrumentation, and Control

1.2.2.7.1 Electrical Power Systems

The plant consists of a single main generator directly connected to a main power transformer through an isolated phase electrical bus duct. The main power transformer steps up the output of the 25-kV generator to a nominal 500-kV transmission system voltage.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-27 The output of the main power transformer is connected to a 500-kV switchyard consisting of circuit breakers, disconnect switches, buses, and associated equipment arranged in a ring bus

configuration.

A 230-kV offsite supply is provided to a separate startup auxiliary transformer to supply

maximum startup, operating and shutdown load requirements for a normal plant auxiliary loads and for safety loads. In addition, a separate 115-kV offsite supply serves a backup auxiliary transformer with sufficient capacity to provide the power requirements of plant safe shutdown

loads. 1.2.2.7.2 Electrical Power Systems Process Control and Instrumentation

Main generator electrical controls are located in the main control room. These include main generator circuit breaker controls, synchronizing equipment, and generator excitation and voltage control equipment. Instrumentation is also provided in the main control room for the

main generator connections and equipment. This includes indicating instruments for voltage, current, kW, MVAR, and frequency. Recording instruments are provided for generator MW output and main bus voltage. Kilowatt-hour meters are provided for main generator outputs and for auxiliary power system loads. Instrumentation is provided for monitoring generator

and transformer temperatures. Other types of monitoring instrumentation are provided as

required to ensure proper operation of equipment. Circuit breaker controls, metering, and indication for the auxiliary power system are also located in the main control room.

High-speed protective relaying equipment is provided for the main generator, main and

auxiliary transformers, main buses, transmission lines, and interconnecting cables and bus ducts to provide proper isolation of this equipment in the event of electrical faults. The

protective relay system includes breaker failure protection and backup relaying to ensure proper isolation of electrical faults in the event of a failure of the primary protective relaying.

1.2.2.7.3 Nuclear System Process Control and Instrumentation

1.2.2.7.3.1 Reactor Manual Control System. The reactor manual control system (RMCS) provides the means by which control rods are positioned from the control room for power

control. The system operates valves in each CRD hydraulic control unit to change control rod position. Only one control rod can be manipulated at a time. The RMCS includes the logic that restricts control rod movement (rod block) under certain conditions as a backup to procedural controls.

1.2.2.7.3.2 Recirculation Flow Control System. During normal power operation, a variable frequency power supply is used to control flow by varying the RRC pump motor speed.

Adjusting the frequency changes motor speed and the coolant flow-rate through the core, thereby changing the core power level.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-28 1.2.2.7.3.3 Neutron Monitoring System. The neutron monitoring system is a system of in-core neutron detectors and out-of-core electronic monitoring equipment. The system provides indication of neutron flux, which can be correlated to thermal power level for the entire range of flux conditions that can exist in the core. The source range monitors (SRM) and the intermediate range monitors (IRM) provide flux level indications during reactor startup and low power operation. The local power range monitors (LPRM) and average power range monitors (APRM) allow assessment of local and overall flux conditions during power range operation. The traversing in-core probe system (TIP) provides a means to calibrate the individual LPRM sensors. The neutron monitoring system provides inputs to the reactor manual control system to initiate rod blocks if preset flux limits are exceeded, and inputs to the RPS to initiate a scram if other limits are exceeded.

1.2.2.7.3.4 Refueling Interlocks. A system of interlocks that restricts movement of refueling equipment and control rods when the reactor is in the refueling and start-up modes is provided to prevent an inadvertent criticality during refueling operations. The interlocks back up procedural controls that have the same objective. The interlocks affect the refueling platform, refueling platform hoists, fuel grapple, and control rods.

1.2.2.7.3.5 Reactor Vessel Instrumentation. In addition to instrumentation for the nuclear safety systems and ESF, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of

the vessel itself. This instrumentation monitors reactor vessel pressure, water level, coolant temperature, reactor core differential pressure, coolant flow rates, and RPV head inner seal

ring leakage.

1.2.2.7.3.6 Process Computer System. An on-line process computer is provided to monitor and log process variables and to make certain analytical computations. The rod worth minimizer function of the computer prevents rod withdrawal under low power conditions if the rod to be withdrawn is not in accordance with a preplanned pattern. The effect of the rod block is to limit the reactivity worth of the control rods by enforcing adherence to the

preplanned rod pattern.

1.2.2.7.4 Power Conversion Systems Process Control and Instrumentation

1.2.2.7.4.1 Digital Electro-Hydraulic Control System. The DEH control system maintains control of the turbine governor valves and turbine bypass valves to allow proper generator and reactor response to system load demand changes while maintaining the nuclear system pressure essentially constant. When the generator is not connected to the grid, the DEH control system maintains turbine-generator speed (frequency) in response to reactor pressure changes by adjusting steam flow to the turbine valves and bypass valves.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-29 The turbine generator speed/load controls can initiate rapid closure of the turbine control (governor) valves and rapid opening of the turbine bypass valves to prevent turbine overspeed

on a generator electric load loss.

1.2.2.7.4.2 Feedwater System Control. A three-element controller is used to regulate the feedwater system so that proper water level is maintained in the reactor vessel. The controller uses main steam flow rate, reactor vessel water level, and feedwater flow rate signals. The feedwater control signal is used to control the speed of the steam turbine-driven feedwater pumps. During startup, shutdown, and low plant load conditions, the steam turbine-driven feedwater pumps are run at constant speed, and the feedwater control signal is used to modulate a startup feedwater control valve to maintain proper reactor water level.

1.2.2.8 Radioactive Waste Systems

1.2.2.8.1 Liquid Radwaste System

This system collects, treats, stores, and disposes of all radioactive liquid wastes. These wastes are accumulated directly in radwaste tanks or in sumps at various locations throughout the plant for subsequent transfer to collection tanks in the radwaste facility. Wastes are processed on a batch basis with each batch being processed by such method or methods appropriate for the quality and quantity of materials determined to be present. Processed liquid wastes may be returned to the condensate system or discharged to the circulating water blowdown line to the river. The liquid wastes in the discharge piping are diluted with circulating water blowdown to achieve a concentration at the site boundary which is below the limits of 10 CFR Part 20.

Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance with minimum personnel exposure. For example, tanks and processing equipment which contain significant radiation sources are located behind shielding, and sumps, pumps, instruments, and valves are located in controlled access rooms or spaces. Processing equipment is selected and designed to require a minimum of maintenance.

Protection against accidental discharge of liquid radioactive waste is provided by design redundancy, instrumentation for detection and alarm of abnormal conditions, and procedural

controls.

1.2.2.8.2 Solid Radwaste System Solid radioactive wastes are collected, processed, and packaged for storage and ultimate burial.

These wastes are generally stored on the site until the short half-lived isotopes have decayed.

Wet solid wastes are collected, dewatered, and solidified in steel containers. Examples of these wastes are filter residue, concentrated wastes , and spent resins. Dry solid wastes such as paper, air filters, rags, and used clothing are compressed and packaged in steel containers.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-30 1.2.2.8.3 Gaseous Radwaste System The purpose of the gaseous radwaste system is to process and control the release of gaseous radioactive wastes to the site environs so that the total radiation exposure to persons outside the controlled area does not exceed the limits of the applicable regulations, 10 CFR 20 and 10 CFR 50, Appendix I, even with some defective fuel rods.

The offgases from the main condenser are the major source of gaseous radioactive waste. The treatment of these gases includes volume reduction through a catalytic hydrogen-oxygen recombiner, water vapor removal through a condenser, decay of short-lived radioisotopes through a holdup line, further condensation, filtration, adsorption of isotopes on activated charcoal beds, further filtration through high efficiency filters, and final release.

Continuous radiation monitors are provided which indicate radioactive release from the reactor and from the charcoal absorbers. The radiation monitors are used to isolate the OG system on high radioactivity to prevent gas of unacceptably high activity from release.

Since clean gland seal steam is used, the offgases from the gland seal steam condenser are not

treated prior to release.

The design of the OG system is such that the annual exposure to any offsite person during normal operation from gaseous sources will be ALARA and less than 10 CFR 20.

1.2.2.9 Radiation Monitoring and Control 1.2.2.9.1 Process Radiation Monitoring Radiation monitors are provided on various lines to monitor either for radioactive materials released to the environs via process liquids and gases or for process system malfunctions. All effluents from the plant which are potentially radioactive are monitored. Several of the effluent monitoring systems record the results prior to discharge as noted on the following list of the major monitoring systems provided.

a. Main steam line radiation monitoring system, b. Air ejector and offgas radiation monitoring systems (results recorded except for the charcoal bed vault), c. Liquid radwaste effluent radiation monitoring system, d. Plant service water and circulating water blowdown radiation monitoring systems, e. Standby service water radiation monitoring system, C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-31 f. Reactor building ventilation exhaust plenum radiation monitoring system (results recorded), g. Reactor building elevated release point radiation monitoring system (results recorded except for particulate/iodine sample), h. Turbine building ventilation exhaust radiation monitoring system, (results recorded), i. Radwaste building ventilation exhaust radiation monitoring system (results recorded), and

j. Reactor building closed cooling water monitoring system.

1.2.2.9.2 Area Radiation Monitors Radiation monitoring devices are provided in key areas throughout the plant buildings to ensure that plant personnel will not be inadvertently exposed to high radiation doses.

1.2.2.9.3 Site Radiological Environmental Monitoring A comprehensive radiation surveillance program was initiated in the spring of 1978 to measure radiation levels in the environs surrounding the plant. The program is designed to measure radiation exposure or radioisotope levels in the environment.

The details of this monitoring program are given in the Offsite Dose Calculation Manual (ODCM). 1.2.2.9.4 Liquid Radwaste System Control Liquid wastes to be discharged are handled on a batch basis with protection against accidental discharge provided by procedural controls. Instrumentation with alarms to detect abnormal concentration of the radwaste is provided, including automatic closure of discharge valves

isolating the system from the environment.

1.2.2.9.5 Solid Radwaste System Control The solid radwaste system collects, treats, and stores solid radioactive wastes for offsite shipment. Wastes are handled on a batch basis. Radiation levels of the various batches are

monitored by the operator.

1.2.2.9.6 Gaseous Radwaste System Control Gaseous radwastes are discharged through a reactor building elevated release point. Radiation levels of the release are continuously monitored and recorded. Isolation of the main condenser C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.2-32 offgas is automatically initiated prior to release should the activity of the offgas exceed discharge limits.

1.2.2.10 Shielding

The shielding in the plant is designed to minimize exposure of plant personnel to radiation.

The radiation levels during operation or shutdown conditions have been considered in determining the shielding requirements.

1.2.2.11 Fuel Handling and Storage Systems 1.2.2.11.1 New and Spent Fuel Storage New and spent fuel storage racks are designed to prevent inadvertent criticality and load buckling. Sufficient coolant and shielding are maintained to prevent overheating and excessive personnel exposure, respectively. The design of the fuel pool provides for corrosion resistance, adherence to Seismic Category I requirements, and prevention of K eff from exceeding 0.95 under dry or flooded conditions.

1.2.2.11.2 Fuel Handling System The fuel handling equipment includes a fuel inspection stand, fuel preparation machine, a 125-ton crane, a refueling platform, a new fuel transfer basket, jib cranes, and other related tools for fuel and reactor servicing.

1.2.2.11.3 Fuel Pool Cooling and Cleanup System

The FPC system removes decay heat from stored spent fuel and maintains specified water

temperature, purity, clarity, and level. This prevents fuel pool boiling and buildup of excessive radioactive materials in the cooling water, thereby minimizing possible exposures to

plant personnel.

Cooling of spent fuel is accomplished by the Seismic Category I cooling system as described in

Section 9.1.3. It can be isolated from the Seismic Category II cleanup portion of the system

by automatic, redundant, Seismic Category I isolation valves which actuate on low fuel pool water level. If required, safety grade cooling and makeup water from the SW system is available to the system by remote-manual operation of redundant Seismic Category I valves to provide long-term cooling and prevent fuel pool boiloff which might result in unacceptable

building environmental conditions.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-03-046 1.2-33 1.2.2.12 Cooling Water and Auxiliary Systems

1.2.2.12.1 Reactor Building Closed Cooling Water System

The RCC system consists of pumps, heat exchangers, controls, and instrumentation to provide adequate cooling for the reactor auxiliary systems. The system is designed to provide a closed cooling water loop between nonessential systems which are potentially radioactive and the

TSW system.

1.2.2.12.2 Plant Service Water System

Normal TSW is supplied from service water pumps located in the circulating water pump house. Two service water pumps are provided. The TSW system is designed to remove heat from various auxiliary equipment located within the plant.

1.2.2.12.3 Ultimate Heat Sink

Two spray ponds that serve as the UHS conservatively have a combined equivalent storage of

30 days, assuming no makeup and maximum evaporation and drift losses. Provisions are made

to replenish the sink to allow continued cooling capability beyond the initial 30-day period.

1.2.2.12.4 Demineralized Water Makeup System

The DW makeup system is comprised of the trailer-mounted demineralizers and the DW system.

The DW system is designed to provide demineralized water to the CSTs for plant makeup and demineralized water for other plant operating requirements.

1.2.2.12.5 Potable Water and Sanitary Drain Systems

The plant potable water (PW) system provides water for drinking and sanitary purposes.

Potable water is normally supplied from the tower makeup system (see Section 9.2.3).

The sanitary drain system effluent is directed to a central sanitary waste treatment facility which uses aerated lagoons in series with lined facultative stabilization ponds. The treatment plant, about 2500 ft SE of the CGS reactor, also receives waste from the WNP-1/4, the Plant

Support Facility, and the DOEs 400 Area.

1.2.2.12.6 Process Sampling Systems

The process sampling system provides process information that is required to monitor plant

and equipment performance and changes to operating parameters. Representative liquid and C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 1.2-34 gas samples are taken automatically and/or manually during normal plant operation for laboratory or on-line analyses.

1.2.2.12.7 Condensate Supply System

The condensate storage facility provides a source of water for testing and makeup during

operation. Two 400,000 gal CSTs are interconnected to simultaneously supply condensate to the main condenser via one header, to the CRD pumps via a second header, and to the RHR, RCIC, and HPCS systems and condensate supply and condensate filter/demineralizer backwash pumps via a third header. The condensate supply pumps deliver condensate to miscellaneous services in the reactor and radwaste buildings.

Condensate is returned to the CSTs from the HPCS, RCIC, and radwaste systems, from CRD, condensate supply, and condensate filter/demineralizer backwash pump mini-flows, and from the main condensate system (equivalent to excess CRD injection water). Initial fill and makeup

is from the DW system.

1.2.2.12.8 Equipment and Floor Drainage Systems

Plant equipment and floor drainage systems handle both radioactive and nonradioactive drains.

Drainage systems which carry radioactive waste are isolated from drainage systems which do

not carry radioactive waste.

All drains in the reactor building and radwaste building are considered radioactive. Turbine building drains are divided into radioactive and nonradioactive but all are directed to the

radwaste system for processing. Floor and equipment drains in the diesel generator building and service building are routed to the storm water drainage system. The storm water drainage system is normally nonradioactive, however some accumulation of radioactive material (notably tritium) can occur.

1.2.2.12.9 Compressed Air Systems

The compressed air system consists of the control and service air system and the containment

instrument air (CIA) system.

The control air system (CAS) is designed to supply clean, dry, oil-free air to station instrumentation and controls and to the accumulators of the MSIVs located outside the primary

containment.

The service air (SA) system is designed to supply clean, oil-free air for station services, such as backwashing demineralizers and filters, hose connections for maintenance throughout the station and breathing air at selected locations.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.2-35 The CIA system is designed to deliver nitrogen or clean, dry, oil-free air for MSIVs, SRV accumulators, and pneumatic operators located inside the primary containment.

1.2.2.12.10 Heating, Ventilating, and Air Conditioning Systems

The HVAC systems are designed to maintain proper air quality for personnel comfort and safety. In addition, the main control room, the critical switchgear area, the cable spreading room HVAC systems, the SW pump room heat removal systems, the reactor building emergency pump and critical electric equipment area cooling systems, and the ventilation system for the standby diesel generators are designed to operate under all station conditions. The primary containment drywell cooling and ventilation system is designed to operate during normal operation and under most upset conditions except a LOCA. All air distribution systems are designed so that airflow is directed from areas of lesser potential contamination to areas of

progressively greater potential contamination.

Three separate and redundant HVAC systems service the main control room, cable spreading room, and critical switchgear areas. SW is used as the cooling medium for each system when

the normal cooling water supply is unavailable.

Heating and ventilation for the standby diesel generator rooms is provided continuously for

each diesel generator unit. Water cooled air handling units provide additional cooling when

the diesel generators operate.

The turbine building is provided with a once-through ventilation system based on the use of

evaporative coolers.

Ventilation for the radwaste building is provided by means of a once-through ventilation system with particulates filtered bef ore release to the atmosphere.

The SW pump room heat removal systems consist of two independent and separate fan coil

units.

The reactor building emergency pump and critical electric equipment area cooling system

consists of 13 air handling units which operate to supply cool air to each of the critical equipment rooms when pumps are started and during abnormal conditions.

The primary containment drywell cooling and ventilation system consists of five fan coil units and nine recirculation fans. During normal operation, a minimum three out of five fan coil

units are operating.

Ventilation for the reactor building is provided by a once-through ventilation system based on the use of evaporative coolers. The system incorporates the necessary isolation valves to C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-04-037 1.2-36 ensure the necessary secondary containment integrity. A drywell and suppression chamber

purge capability is provided as part of this system.

Other HVAC systems provide ventilation to the service building and other miscellaneous areas.

1.2.2.12.11 Fire Protection System

The FP system is designed to provide for the detection and extinguishing of fires.

Manual pull stations and automatic fire detectors are located appropriately throughout the plant and fire alarms are annunciated in the main control room.

The FP system provides a reliable water distribution system for extinguishing fires. Two motor-driven fire pumps are used for normal service, with a diesel-engine-driven fire pump as a backup. A second diesel-driven fire pump with a separate water supply provides an additional backup. Motor-driven jockey pump is provided to maintain system pressure and to prevent cycling of the main fire pumps.

Automatic suppression systems provide protection to higher hazard areas of the plant including:

Deluge systems protect the transformers and most other areas containing oil piping and oil storage equipment.

A low-pressure carbon dioxide (CO

2) system is provided for the generator exciter housing. A total flooding Halon system is provided for the main control room power generation control complex (PGCC) subfloor.

Wet pipe sprinklers protect the turbine/generator bearings and other miscellaneous areas. Preaction sprinkler systems protect diesel generators, day tank/transfer pump rooms, and areas with high concentrations of electrical cables.

Manual suppression includes:

Fire hydrants spaced around the yard fire main loop.

Fire hose stations located throughout the plant.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-04-037 1.2-37 Portable fire extinguishers of appropriate types are strategically and conspicuously placed throughout the plant.

1.2.2.12.12 Communications Systems

The plant communication systems are designed to provide reliable communication inside and outside the plant and from the plant to local fire protection and law enforcement authorities.

The system utilizes a public address and building wide alarm system, a public telephone system, a private digital telephone system, a sound powered telephone system, a radio communication system, and an automatic transmission telephone link to the Dittmer Control Center of the Bonneville Power Administration (BPA).

1.2.2.12.13 Lighting Systems

The plant lighting systems are normal ac lighting, normal-emergency ac lighting, dc lighting, and battery-pack emergency lighting. Lighting intensities are designed to provide indoor and outdoor illumination consistent with the July 1974 Illumination Engineering Society

recommendations, and meet or exceed Occupational Safety and Health Act (OSHA)

requirements.

1.2.2.12.14 Normal Auxiliary Alternating Current Power System

The plant normal auxiliary ac power system consists of two normal auxiliary transformers, the 4.16-kV and 6.9-kV normal auxiliary (non-Class 1E) distribution system, the 480-V auxiliary

power distribution system and the 120/208-V non-Class 1E distribution system.

The normal ac auxiliary transformers provide power to all plant auxiliaries and comprise the normal plant ac power source when the main generator is operating. One of the normal

auxiliary transformers is a dual secondary type with both secondary windings stepping down the generator voltage to 4.16 kV for supply to 4.16-kV non-Class 1E switchgear buses. The other normal auxiliary transformer steps down the generator voltage to 6.9 kV for supply of

6.9-kV non-Class 1E switchgear buses.

The plant 480-V ac auxiliary power system distributes ac power necessary for normal auxiliary

and ESF 480-V plant loads. All non-ESF elements of this distribution system are capable of being supplied from the normal auxiliary power source or from the startup power source via the 4.16 kV-non-Class 1E switchgear. The ESF portions of the 480-V distribution system are supplied via the 4.16-kV Class 1E switchgear, and therefore are capable of being supplied by either the normal, startup, backup, or standby sources.

The 120/208-V non-Class 1E ac power system provides power for non-ESF loads.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-019 1.2-38 1.2.2.12.15 Diesel Generator Fuel-Oil Storage and Transfer System

The diesel fuel oil storage and transfer system consists of separate, independent diesel oil supply subsystems serving each of two emergency diesel generators and the HPCS diesel

generator. Each full capacity subsystem consists of a fuel oil storage tank, a transfer pump, a

day tank, interconnecting piping, strainers and valves, and associated instrumentation and

controls.

1.2.2.12.16 Auxiliary Steam System The auxiliary steam (AS) system normally operates only when the heating steam evaporators

are inoperative during plant shutdown. The system then supplies steam to HVAC systems for air and water space heating and for humidification and also to the radwaste system. The system consists of fuel oil storage tank and transfer pumps, auxiliary boiler, blowdown tank, chemical feed tank and metering pump, deaerator and boiler feed pumps, condensate return tank pumps, steam supply and condensate return piping and valves, and associated instruments

and controls.

1.2.3 COMPLIANCE

WITH NRC REGULATORY GUIDES

The CGS conformance to the NRC regulatory guides is documented in Section 1.8 and in appropriate sections of this FSAR.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.2-39 Table 1.2-1 Principal Regulations and Codes Followed in Plant Design Number Title 10 CFR series Code of Federal Regulations, principally:

10 CFR 20 Standards for Protection Against Radiation 10 CFR 50 Licensing of Production and Utilization Facilities 10 CFR 50, Appendix A General Design Criteria for Nuclear Power Plant Construction Permits 10 CFR 50, Appendix B Quality Assurance Criteria 10 CFR 50, Appendix I Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low As Is Reasonably Achievable" 10 CFR 100 Reactor Site Criteria IEEE-279 IEEE Criteria for Nuclear Power Generating Station Protection Systems IEEE-308 IEEE Criteria for Class IE Electrical Systems for Nuclear Power Generating Stations ASME B&PV ASME Boiler and Pressure Vessel Code: Section III Nuclear Components Section VIII Pressure VesselsSection XI Inservice Inspection

AEC Press Release IN-817 Tentative Regulatory Supplementary Criteria for ASME

Code-Constructed Pressure Vessels ANSI-B31.1.0 ANSI Standard Code for Pressure Piping, Power Piping

NOTE: Additional codes and regulations applying to specific areas of system design are

referenced in discussions of individual systems.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.2-40 Table 1.2-2 Plant Shielding and Zone Classification Zone Description Design Dose Rate (mrem/hr) I Uncontrolled, unlimited access 1.0 II Controlled, limited access 2.5 III Controlled, occupancy for short periods, normally inaccessible 100 IV For very short periods. Secured and controlled entrance.

>100 NOTES:

1. Radiation Zone I areas can be occupied by plant personnel or visitors for unlimited periods.

2. Radiation Zone II areas are areas where whole body dose is not expected to exceed 1.25 rem per calendar quarter.

3. Areas having dose rates in excess of 100 mrem/hr are posted as high radiation areas and access is secured and controlled.

4. Radiation Zone III and IV areas can be entered only after the radiation level is determined and the working time limit is established.

5. Accessible areas have dose rates of less than 100 mrem/hr.

6. Access to all controlled areas is through controlled check points.

7. Controlled and limited access areas are identified by warning signs.

Equipment AcronymsAAAudio AlarmACAir Conditioning UnitACCAccumulator ACMAcoustic Monitor/SensorADAir Damper AHAir Handling Unit

AI Air IndicatorALMAlarm Annunciator-Do Not Use ALTAlternating Relay AMAmmeter AMPAmplifier ANNAnnunciator AOAir Operator ARAir Receiver AR/FRAnalyzer and Flow Recorder ASMAssembly ASWAir Switch (4-way Valve)

ATAir Transmitter ATDAmp Transducer ATSAutomatic Transfer Switch AUDAudio Monitor AUXAuxiliary Unit AVAir Valve AW Air Washer AYAnalyzer B024 Volt Battery B1125 Volt Battery B2250 Volt Battery B312 Volt Battery B448 Volt Battery BDETBadge (Keycard) Detector BELLBell (Fire Protection)

BFIBlown Fuse Indicator BLBaler BLDGBldg (For PSD System Only)

BLRBoiler BTBolted Tee (For SA System)

BUEmerg Lighting Battery Unit BUOYBuoy C CompressorC024 Volt Battery Charger C1125 Volt Battery Charger C2250 Volt Battery Charger C312 Volt Battery Charger CABCabinet CAPCapacitor CBCircuit Breaker CCCooling Coil CCTVClosed Circuit Television CCUCentral Control Unit CEConductivity Element CERACond Element Retractor Assembly CFCharcoal Filter CFGCentrifuge CHChannel CHLChlorinators CHMChamber CHRChiller CHSChassis CIConductivity Indicator CICConductivity Ind Controller CISConductivity Ind Switch CITConductivity Ind TransmitterCITSConductivity Ind Transmitter SwitchCJWCooling Jacket Water CMCommunications Monitor CNTRContractor COECorrosivity Element COICCorrosivity Indic Cont COMPComputer CONNConnector CORCorrosivity Recorder COSCarbon Monoxide Sensor COTCorrosivity TransmitterCPControl Panel CPLData Coupler CPTRCompactorCPUCentral Processing Unit CRConductivity Recorder; Control Room ChillerCRACraneCRBControl Rod Blade CRMControl Module CRSConductivity Recorder Switch CRTTerminal Display Screen CSConductivity Switch CSKShield Transfer CaskCTCurrent Transformer/Cooling Tower CUCondensing Unit

D Damper (Backdraft Or Motor)DCDecoder DCMDry Cleaning Machine DCNCRD Decontamination System DDRDisk Drive Recorder DEDensity Element DETDetector DFSDifferential Flow Switch DGDigital Display Generator DHDrywell Head DIFDiffuser DIODiode, Control Rectifier DISCDisconnect Switch DLRDifferential Level Recorder DLSDifferential Level Switch DLTDifferential Level Transmitter DMDemineralizer DMMDisplay Memory Module DMSDemister DMTRDemand Meter DOEDissolved Oxygen Element DOITDissolved Oxygen Indic Trans DOORDoor DORDissolved Oxygen Recorder DPDistribution Panel DPCDiff Press Controller DPEDrip Pan Elbow DPIDiff Press Ind DPICDiff Press Ind Controller DPIRDiff Press Ind Recorder DPISDiff Press Ind Switch DPITDiff Press Ind Transmitter DPRDiff Press Recorder DPSDiff Press Switch DPTDiff Press Transmitter DRDemand RecorderDRVEDrive Mechanism For CRDDSDensity Switch DT Dens Trans Or Drive T urbineDTISDiff Temp Indicating SwitchDTRSDiff Temp Recording Switch DTSDiff Temp Switch DTTDiff Temp Transmitter DUDeaerator DVDeluge Valve DVSPDump Valve Solenoid Pilot DVSPVDump Valve Solenoid Pilot ValveDWSDemineralized Water Shower DYDryer E/IVolt To Current Converter E/PElectro Pneumatic Converter E/SElectronic Power Supply EAMPPreamplifier ECElectronic Controller ECGElectrochemical Generator EDEductor EFElectronic Filter EFCExcess Flow Check Valve EHCElectric Heating Coil EHOElectrohydraulic Operator

EI Power Supply MonitorEISPower Supply Monitor Switch EJExpansion Joint EJCEjector ELEVElevator ELPEmergency Lighting Panel EMSQMean Square Voltage Device ENGEngine EPAElectrical Protection Assem EPPEmergency Power PanelEQSpeciality Equip and ToolsERBEmerg Rmt Ballast (Lighting)

ESExhaust Silencer ESHElectric Strip Heater EUHElectric Unit Heater EVEvaporator EXExhauster EXCExciter

F FilterF/UFlow Unit FAFlame Arrestor FCFlow Controller FCNFuel Oil Tk Fill Connector FCVFlow Control Valve FDFire Damper FDgFreon Degreaser FEFlow Element FGFlow Glass FGENFunction Generator FHFume Hood FHBFuel Handling Box

FI Flow IndicatorFICFlow Indicating Controller FICSFlow Indicating Controller Switch FISFlow Indicating Switch FITFlow Indicating Transmitter FLFilter FLPFillport Assem FLTFilter FLXFlexible Connection FNFan FOFreon Actuated Operator FPFilter Polisher FQFlow Integrator FQIFlow Integrating Indicator FQSFlow Integrating Switch FRFlow Recorder FR/DLFlow and Diff. Level Recorder FRCFlow Recording Controller FRDLRFlow and Diff Level Recorder FRSFlow Recording Switch FSFlow Switch FSPVFlow Solenoid Pilot Valve FTFlow Transmitter FTDFrequency TransducerFUFilter Unit FUSEFuse FXFlow Test Connection FYFlow Sig. Cond.

GATEGateGCALAGS CalibratorGENGenerator GOVGovernor GVTGravity Ventilator

H Heater H 2EHydrogen Element H 2IHydrogen Indicator H 2ISH 2 Indicating Switch/Monitor H 2ITHydrogen Ind Transmitter H 2RHydrogen Recorder H 2THydrogen TransmitterHASHigh Amplitude SelectorHCHeating Coil HCUHydraulic Control Unit HFHEPA Filter HMHour MeterHOHydraulic OperatorHOIHoist HPValve Act. Hyd. Power Unit HPUHydraulic Power Unit HRHydrogen Recombiner HSHose Station HSSHigh Selector Switch HTHydrant HTCHeat Trace Cable HTPHeat Trace Panel HUHumidifier HUMHumidifier (Obsolete. Use HU)

HVHeating and Ventilation Unit HVRBHigh Voltage Rubber Blanket HXHeat ExchangerHZMHertz Meter I/PCurrent Pneumatic Converter IDIonization Detector

IL Indicating LightIMDInductive Motor Drive INInverter INDInductor INDXIndexer IOSCurrent Operated Switch IRInstrument Rack

IS Intake SilencerISOLIsolator, Isolation Device ITDCurrent Transducer

IX Ion ExchangerJBJunction Box JPJet Pump KBDComputer Keyboard (Security)

L LubricatorLALightning Arrestor LAGDynamic Compensator LASLow Amplitude Selector LCLevel Controller

LCRM Log Count Rate MeterLCVLevel Control Valve LELevel Element LFLighting Fixture LGLevel Glass

LI Level IndicatorLICLevel Indicating Controller LISLevel Indicating Switch LITSLevel Indic Trans SwitchLMSLimit Switch LMTRV/I Signal Limiter LNRLinear Reactor LOCLube Oil Conditioner LPLighting Panel LPW24 Volt Lambda Power SupplyLRLevel Recorder LR/PRLevel/Pressure Recorder LRSLevel Recording Switch LSLevel Switch LSCLightning Strike Counter LSPVSol. Pilot Valve TMU-level LSSLow Selector Switch LTLevel TransmitterLTDLevel Transmitter DetectorLVDTLinear Var. Dif. Transformer LVSLow Volume SelectorLWSLow Differential Pressure M MotorM/AManual/Auto Station MAManifold MACHMachine MBSMaint. Bypass Switchgear MCMoisture Controller MDETMetal Detector MDSManual Discharge Station MDUMotion Detection Unit MEMoisture Element MGMotor-Generator Set MHDDMoving Head Disc Drive MIMoisture Indicator MICMoisture Indicating Controller MISMoisture Indicating Switch MMMotor Module (TIP System)

MOMotor Operator MODEMModem MONMonitor MPDSMicroprocessor Data System MPSManual Pull Station MRMoisture Recorder MSMoisture Separator MTMositure Transmitter MTADew Point Transmitter AmplifMTSManual Transfer SwitchMUXMultiplexer MVManifold Valve MV/IM/Volt To Current Converter MWMicrowave Receiver MXMixerHMIHuman Machine Interface

Input Output Aux Input Signal Identification or MPL No.Signal Present

when Permissive This block defines a permissive function which must be satisfied to permit the signal flow to pass to the next block. This block has incoming, outgoing,and may have auxiliary signals. The output from this permissive may be sealed in.

Permissive Device Location Number

or LP CR Local Sheet & Zone

Later Note: The word later may be used if the location is unknown but the correct location shall be

noted on a future revision.

= Local Panel

= Control Room

= Mounted Locally

= See Drawing Sheet

= See Note

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-1 1.3 COMPARISON TABLES The italicized information is historical and was provided to support the application for an

operating license.

1.3.1 COMPARISONS

WITH SIMILAR FACILITY DESIGNS

This section highlights the principal design features of CGS and compares its major features with other boiling water reactor (BWR) facilities. The design of this facility is based on proven technology obtained during the development, design, construction, and operation of BWRs of similar types. The data, performance, characteristics, and other information presented here represent the design of the facilities at the time of the CGS operating license review.

1.3.1.1 Nuclear Steam Supply System Design Characteristics

Table 1.3-1 summarizes the design and operating characteristics for the nuclear steam supply systems. Parameters are related to rated power output for a single plant unless otherwise noted. The fuel thermal, hydraulic, and nuclear design data are that for the initial core load.

Cycle specific data are provided in Chapter 4 , Section 5.2 , and Appendix 15F.

1.3.1.2 Power Conversion System Design Characteristics

Table 1.3-2 compares the power conversion system design characteristics.

1.3.1.3 Engineered Safety Features Design Characteristics

Table 1.3-3 compares the engineered safety features design characteristics.

1.3.1.4 Containment Design Characteristics

Table 1.3-4 compares the containment design characteristics.

1.3.1.5 Radioactive Waste Management Systems Design Characteristics

Table 1.3-5 compares the radioactive waste management design characteristics.

1.3.1.6 Structural Design Characteristics

Table 1.3-6 compares the structural design characteristics.

1.3.1.7 Electrical Power Systems Design Characteristics

Table 1.3-7 compares the electrical power systems design characteristics.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-2 1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION

Significant changes that have been made in the facility design since submission of the PSAR are

listed in Table 1.3-8. Items in Table 1.3-8 are cross referenced to the appropriate portion of the FSAR that describes the changes and the bases for them.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-3 Table 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics a CGS b BWR 5 251-764 HATCH 1 c BWR 4 218-560 ZIMMER c BWR 5 218-560 Thermal and Hydraulic Design (see Section 4.4) Rated power (MWt) 3323 2436 2436 Design power (MWt) (ECCS design basis) 3468 2550 2550 Steam flow rate (1b/hr) 14.295 x 10 6 10.03 x 10 6 10.477 x 10 6 Core coolant flow rate (1b/hr) 108.5 x 10 6 78.5 x 10 6 78.5 x 10 6 Feedwater flow rate (1b/hr) 14.256 x 10 6 10.445 x 10 6 10.477 x 10 6 System pressure, nominal in steam dome (psia) 1020 1020 1020 Average power density (KW/liter) 49.15 51.2 50.51 Maximum thermal output (KW/ft) 13.4 13.4 13.4 Average thermal output (KW/ft) 5.38 7.11 5.45 Maximum heat flux (Btu/hr-ft

2) 428,360 428,300 354,000 Average heat flux (Btu/hr-ft

2) 145,060 164,700 143,900 Maximum UO 2 temperature (°F) 4380 4380 3325 Average volumetric fuel temperature (°F) 1100 1100 1100 Average cladding surface temperature (°F) 558 558 558 Minimum critical power ratio (MCPR) 1.24 1.9 d 1.21 Coolant enthalpy at core inlet (Btu/1b) 527.6 526.2 527.4 Core maximum exit voids within assemblies 79 79 75 Core average exit quality (% steam) 13.5 12.9 13.6 Feedwater temperature (°F) 420 387.4 420 Design power peaking factor Maximum relative assembly power 1.40 1.40 1.40 Local peaking factor 1.15 1.24 1.24 Axial peaking factor 1.40 1.5 1.4 Total peaking factor 2.51 2.6 2.43 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-4 Table 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics a (Continued)

CGS b BWR 5 251-764 HATCH 1 c BWR 4 218-560 ZIMMER c BWR 5 218-560 Nuclear Design (First Core)

(see Section 4.3) Water/UO 2 volume ratio (cold) 2.55 2.53 2.41 Reactivity with strongest control rod out (k eff) <0.99 <0.99 <0.99 Moderator void coefficient Hot, no voids (k/k - %void)

-1.0 x 10-3 -1.0 x 10 1.0 x 10

-3 At rated output (k/k - %void)

-1.6 x 10-3 -1.6 x 10

-3 1.6 x 10-3 Fuel temperature doppler coefficient At 68°F (k/k - °F fuel)

-1.3 x 10-5 -1.3 x 10 1.3 x 10

-5 Hot, no voids (k/k - °F fuel)

-1.2 x 10-5 -1.2 x 10 1.2 x 10

-5 At rated output (k/k - °F fuel)

-1.3 x 10-5 -1.3 x 10 1.3 x 10

-5 Initial average 235 U enrichment wt (%)

1.91 2.23 1.90 Fuel average discharge exposure (MWd/short ton) 10,300 19,000 15,053 Core Mechanical Design (see Sections 4.2 and 7.6) Fuel assembly Number of fuel assemblies 764 560 560 Fuel rod array 8 x 8 7 x 7 8 x 8 Overall dimensions (in.)

176 176 176 Weight of UO 2 per assembly (1b) (pellet type) 458 (chamfered)490.4 (undished) 483.4 (dished) 465.15 Weight of fuel assembly (1b) 600 681 (undished) 675 (dished) 698 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-5 Table 1.3-1 Comparison Of Nuclear Steam Supply System Design Characteristics a (Continued)

CGS b BWR 5 251-764 HATCH 1 c BWR 4 218-560 ZIMMER c BWR 5 218-560 Core Mechanical Design (see Sections 4.2 and 7.6) (Continued)

Fuel rods (NEDE-20944P) Number per fuel assembly 62 49 63 Outside diameter (in.) 0.483 0.563 0.493 Cladding thickness (in.) 0.032 0.032 0.034 Cap. pellet to cladding (in.) 0.0045 0.006 0.0045 Length of gas plenum (in.)

10 16 14 Cladding material e Zircaloy-2 Zircaloy-2 Zircaloy-2 Fuel pellets Material UO 2 UO 2 UO 2 Density (% of theoretical) 95 95 95 Diameter (in.) 0.410 0.487 0.416 Length (in.) 0.410 0.5 0.420 Fuel channel Overall dimension, length (in.) 166.9 166.9 166.9 Thickness (in.) 0.100 0.080 0.100 Cross section dimensions (in.) 5.494 x 5.494 5.44 x 5.44 5.48 x 5.48 Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Core assembly Fuel weight as UO 2 (1b) 349,900 272,850 260,538 Core diameter (equivalent) (in.) 187.1 160.2 160.2 Core height (active fuel) (in.) 150 144 146 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-6 Table 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics a (Continued)

CGS b BWR 5 251-764 HATCH 1 c BWR 4 218-560 ZIMMER c BWR 5 218-560 Core Mechanical Design (see Sections 4.2 and 7.6) (Continued)

Reactor control system Method of variation of reactor power Movable control rods and variable forced coolant flow Movable control

rods and variable forced

coolant flow Movable control

rods and variable forced coolant flow Number of movable control rods 185 137 137 Shape of movable control rods Cruciform Cruciform Cruciforn Pitch of movable control rods 12.0 12.0 12.0 Control material in movable rods B 4 C granules compacted in SS

tubes B 4 C granules compacted in SS

tubes B 4 C granules compacted in SS

tubes Type of control rod drives Bottom entry locking piston Bottom entry

locking piston Bottom entry

locking piston Type of temporary reactivity control for initial core Burnable poison; gadoliniaurania

fuel rods Burnable poison; gadoliniaurania

fuel rods Burnable poison; gadoliniaurania

fuel rods In-core neutron instrumentation Number of in-core neutron detectors (fixed) 172 124 124 Number of in-core detector assemblies 43 31 31 Number of detectors per assembly 4 4 4 Number of flux mapping neutron detectors 5 4 4 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-7 Table 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics a (Continued)

CGS b BWR 5 251-764 HATCH 1 c BWR 4 218-560 ZIMMER c BWR 5 218-560 Core Mechanical Design (see Sections 4.2 and 7.6) (Continued)

In-core neutron instrumentation (Continued) Range (and number) of detectors Source range monitor Source to 0.001% power (4)f Source to 0.001% power

(4)f Source to 0.001% power

(4)f Intermediate range monitor 0.001% to 10% power

(8)f 0.001% to

10% power (8) f 0.001% to

10% power (8) f Local power range monitor 5% to 125%

power (172) f 5% to 125%

power (124) f 5% to 125%

power (124) f Average power range monitor 2.5% to 125%

power (6)f 2.5% to 125%

power (6)f 2.5% to 125%

power (6)f Number and type of in-core neutron sources 7 Sb-Be 5 Sb-Be 5 Sb-Be Reactor Vessel Design (see Section 5.3) Material Carbon steel stainless clad Carbon steel

stainless clad Carbon steel

stainless clad Design pressure (psig) 1250 1265 1250 Design temperature (°F) 575 575 575 Inside diameter (ft-in.) 20-11 18-2 18-2 Inside height (ft-in.) 72-11 69-4 69-4 Minimum base metal thickness (cylindrical

section) (in.) 6.75 5.53 5.375 Minimum cladding thickness (in.)

1/8 1/8 1/8 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-8 Table 1.3-1 Comparison Of Nuclear Steam Supply System Design Characteristics a (Continued)

CGS b BWR 5 251-764 HATCH 1 c BWR 4 218-560 ZIMMER c BWR 5 218-560 Reactor Coolant Recirculation Design (see Sections 5.1 , 5.2 , and 5.4) Number of recirculation loops 2 2 2 Design pressure: Inlet leg (psig) 1250 1148 1250 Outlet leg (psig) 1650;g 1550 h 1274 1675; g 1575 h Design temperature (°F) 575 562 575 Pipe diameter (in.)

24 28 20 Pipe material (ANSI) 304/316 304/316 304/316 Recirculation pump flow rate (gpm) 47,200 42,200 33,880 Number of jet pumps in reactor 20 20 20 Main Steam lines (see Section 5.4) Number of steam lines 4 4 4 Design pressure (psig) 1250 1146 1250 Design temperature (°F) 575 563 575 Pipe diameter (in.)

26 24 24 Pipe material Carbon steel Carbon steel Carbon steel a Parameters are related to rated power output for a single plant unless otherwise noted.

b See Section

1.3.1 regarding

the status of the data presented here.

c Values correspond to original licensing.

d For Hatch, minimum critical heat flux ratio (MCHFR) was used.

e Free-standing loaded tubes.

f Channels of monitors from LPRM detectors.

g Pump and discharge piping to and including discharge block valve.

h Discharge piping from discharge block valve to vessel.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-9 Table 1.3-2 Comparison of Power Conversion System Design Characteristics

CGS BWR 5 251-764 HATCH I a BWR 4 218-560 ZIMMER a BWR 5 218-560 Turbine Generator (see Sections 10.2 and 10.4) Rated power (MWt) 3468 b 2550 2550 Rated power (MWe) (gross) 1205 b 813 883 Generator Speed (rpm) 1800 1800 1800 Rated steam flow (1b/hr) 15.018 x 10 6b 10.48 x 10 6 11.0 x 10 6 Inlet pressure (psia) 955 950 950 Steam Bypass System (see Section 10.4.4) Capacity (% design steam flow) 25 25 25 Main Condenser (see Section 10.4.1) Heat removal capacity (Btu/hr) 7702 x 10 6 5720 x 10 6 7053 x 10 6 Circulating Water System (see Section 10.4.5) Number of pumps 3 2 3 Flow rate (gpm/pump) 186,000 185,000 150,000 Condensate and Feedwater System (see Section 10.4.7) Design flow rate (1b/hr) 14.26 x 10 6 10.096 x 10 6 10.971 x 10 6 Number of condensate pumps 3 3 3 Number of condensate booster pumps 3 3 3 Number of feedwater pumps 2 2 2 Number of feedwater booster pumps None None None Condensate pump drive ac power ac power ac power Booster pump drive ac power ac power ac power Feedwater pump drive Turbine Turbine Turbine a Values correspond to original licensing.

b Maximum calculated value.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-10 Table 1.3-3 Comparison of Engineered Safety Features Design Characteristics CGS BWR 5 251-764 HATCH I BWR 4 218-560 ZIMMER BWR 5 218-560 Emergency Core Cooling Systems (systems sized on design power) (see Section 6.3) Low pressure core spray systems Number of loops 1 2 1 Flow rate (gpm) 6350 at 128 psid 4625 at 120 psid 4725 at 119 psid High pressure core spray system Number of loops 1 1 a 1 Flow rate (gpm) 1550 at 1130 psid 4250 1330 at 1110 psid 6350 at 200 psid 4725 at 200 psid Automatic depressurization system Number of relief valves 7 7 7 Low pressure coolant injection b Number of loops 3 2 3 Number of pumps 3 4 3 Flow rate (gpm/pump) 7450 at 26 psid 7700 at 20 psid 5050 at 20 psid Residual Heat Removal System (see Section 5.4.7) Reactor shutdown cooling mode: Number of loops 2 2 2 Number of pumps 2 4 2 Flow rate (gpm/pump) c 7450 7700 5050 Duty (Btu/hr/heat exchanger) d 41.6 x 10 6 32 x 10 6 30.8 x 10 6 Number of heat exchangers 2 2 2 Primary containment cooling mode: Flow rate (gpm) 7450 e 30,800 5050 e

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-11 Table 1.3-3 Comparison of Engineered Safety Features Design Characteristics (Continued)

CGS BWR 5 251-764 HATCH I BWR 4 218-560 ZIMMER BWR 5 218-560 Standby Service Water System (see Section 9.2.7) Flow rate (gpm/heat exchanger) 7400 8000 5000 Number of pumps 3 f 4 4 Reactor Core Isolation Cooling System (see Section 5.4.6) Flow rate (gpm) 600 at 1150 psid 400 at 1120 psid 400 at 1120 psid Fuel Pool Cooling and Cleanup System (see Section 9.1.3) Capacity (Btu/hr) 8.0 x 10 6 5.7 x 10 6 6.6 x 10 6 a High-pressure coolant injection system utilized.

b A mode of RHR system.

c Capacity during reactor flooding mode with more than one pump running.

d Heat exchanger duty at 20 hr following reactor shutdown.

e Flow per heat exchanger.

f Includes HPCS service water pumps.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-12 Table 1.3-4 Comparison of Containment Design Characteristics CGS BWR 5 251-764 HATCH 1 BWR 4 218-560 ZIMMER BWR 5 218-560 Primary Containment a (see Sections 3.8.2 and 6.2.2) Type Over and under pressure suppression Pressure suppression Over and under pressure suppression Construction Steel-free standing Steel-free

standing Concrete pre-

stressed with

steel liner Drywell Frustum of cone upper portion Light bulb/steel vessel Frustum of cone

upper portion Pressure-suppression chamber Cylindrical lower portion

with eliptical

bottom Torus/steel vessel Cylindrical

lower portion Pressure-suppression chamber internal

design pressure (psig) 45 56 45 Pressure-suppression chamber external design pressure (psi) 2 2 2 Drywell internal design pressure (psig) 45 56 45 Drywell external design pressure (psi) 2 2 2 Drywell free volume (ft

3) 200,540 b 146,240 180,000 Pressure-suppression chamber free volume (ft 3) 144,184 max 110,950 93,000 Pressure-suppression pool water volume (ft 3) 112,197 min c 87,300 102,000 Submergence of downcomer vent pipe below pressure pool surface (ft) 12 max. 11.67 min.

3.67 10 Design temperature of drywell (°F) 340 281 340 Design temperature of pressure-suppression chamber (°F) 275 281 275 Downcomer vent pipe pressure loss factor 1.9 6.21 2.17 Break area/total vent area 0.105 0.0194 0.008 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-13 Table 1.3-4 Comparison of Containment Design Characteristics (Continued)

CGS BWR 5 251-764 HATCH 1 BWR 4 218-560 ZIMMER BWR 5 218-560 Primary Containment a (see Sections 3.8.2 and 6.2.2) (Continued)

Calculated maximum pressure after blowdown to dwell (no pre-surge) (psig) 34.7 46.5 40.4 Pressure-suppression chamber (psig) 27.6 28 35.6 Initial pressure-suppression pool temperature

rise (°F) 35 50 35 Leakage rate (% free volume/day at 45 psig and 200°F) 0.5 1.2 at 59 psig 0.635 Secondary Containment (see Sections 3.8.4 and 6.2.3) Type Controlled

leakage, elevated release Controlled

leakage, elevated release Controlled

leakage, elevated release Construction Lower levels Reinforced concrete Reinforced concrete Reinforced

concrete Upper levels Steel super-structure and

siding Steel super-

structure and

siding Steel super-

structure and

siding Roof Steel decking Steel decking Steel decking Internal negative design pressure (in. H 2 O) 0.25 0.25 0.25 Design inleakage rate (% free volume/day at 0.25 in. H 2 O) 100 100 100 a Where applicable, containment parameters are based on design power.

b Maximum water level in suppression pool.

c Does not include the water within the reactor pedestal (10,065 ft

3) or the 12 ft of water below the downcomer vent pipe exits (15,000 ft 3).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-14 Table 1.3-5 Radioactive Waste Management Systems Design Characteristics CGS BWR 5 251-764 HATCH 1 BWR 4 218-560 ZIMMER BWR 5 218-560 Gaseous Radwaste (see Section 11.3) Design Bases (noble gases Ci/sec) 100,000 at 30 minutes 100,000 at 30 minutes 100,000 at 30 minutes Process treatment Low temperature

charcoal Recombiner

ambient charcoal Chilled charcoalNumber of beds 8 12 5 Design condenser in-leakage (cfm) 30 40 12.5 Release point - height above ground (ft) 230 394 172 Liquid Radwaste (see Section 11.2) Treatment of

1. Floor drains a F, D, and R F, D, and R F, E, and R 2. Equipment drains a F, D, and R F, D, and R F, D, and R

3. Chemical drains a Neutralized, E, D, and R F, discharged E, solid to

radwaste E, D, concentrates to

solid radwaste

distillate R

4. Detergent drains a Chemical addition, F, E, and sent to

circulating

water discharge b Diluted and sent to circulating

water discharge Reverse osmosis discharge 5. Expected annual average release (Ci) (excluding tritium) 170 2000 1.09 a Legend: D = demineralized. F = filtered.

E = evaporator/concentrator.

R = recycled, i.e., returned to condensate storage.

b Laundry will be processed offsite by authorized contractor.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-15 Table 1.3-6 Comparison of Structural Design Characteristics CGS BWR 5 251-764 HATCH 1 BWR 4 218-560 ZIMMER BWR 5 218-560 Seismic Design (see Section 3.7) Operating basis earthquake (horizontal g) 0.125 0.08 0.10 Safe shutdown earthquake (horizontal g) 0.250 0.15 0.20 Wind Design (see Section 3.3) Maximum sustained (mph) 100 105 90 Tornados Translational (mph) 60 60 60 Tangential (mph) 300 300 300 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-16 Table 1.3-7 Comparison of Electrical Systems Design Characteristics CGS a BWR 5 251-764 HATCH 1 BWR 4 218-560 ZIMMER BWR 5 218-560 Transmission System (see Section 8.2) Outgoing lines (number - rating) 1 - 500 kV 4 - 230 kV 3 - 345 kV Normal auxillary ac power Incoming lines (number - rating) 1 - 230 kV 1 - 115 kV 4 - 230 kV 1 - 69 kV 1 - 345 kV Normal auxiliary transformers 2 2 1 (unit auxiliary) Startup/backup auxiliary transformers 2 2 2 Standby ac power supply Number of diesel generators 3 b 3 c 3 Number of 4160-V shutdown (Class 1E) buses 3 b 3 3 Number of 480-V shutdown (Class 1E) buses 5 b 2 (600 V) 5 Power Supply (dc) (see Section 8.3.2) Number of 24-V batteries 4 2 (48 V)

Number of 125-V batteries 6 d 3 3 Number of 250-V batteries 1 2 1 Number of 24-V buses 2 2 (24/48 V)

Number of 125-V buses 6 d 3 3 Number of 250-V buses 1 2 1 a Does not include 450-V dc security system.

b HPCS system included.

c Total of five for two units.

d HPCS battery and bus included.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000LDCN-99-000 1.3-17 Table 1.3-8 Significant Design Changes from PSAR to FSAR Item Change Reason for Change FSAR Portion in Which Change is Discussed Offgas system class

change The offgas system components are Quality Group C, whereas the system components were described in the PSAR as being Quality Group D. Improve assurance of system integrity.

11.3.1 Control rod drive

position indication Changed to 11 wire probe and solid state. Improved reliability and increased frequency of checking actual rod position.

7.7.1 Control

rod drive system Deleted CRD return line and pump test bypass, revised

cooling and exhaust water headers, added relief valves interconnecting cooling water and exhaust headers, redirected system exhaust flow through the multiple solenoid valves in each HCU. GE recommendation. 4.6.1.1.2.4 Recirculation pump

and motor The flow rate and horsepower required has been reduced; voltage has changed from 4160 V to 6600 V.

A low-frequency motor generator set was added to provide 25% speed.

Detailed system.

5.4.1 Jet pumps The jet pump design was changed to improve five-hole type. Design improvement, increased efficiency. - Recirculation flow measurement The recirculation flow measurement design was changed from a flow element to an elbow-tap type. To improve flow measurement accuracy.

7.3.1 Recirculation

system The pressure interlock for RHR injection was changed. IEEE-279 requirements. 7.3.1, 7.6.1 Recirculation system Bypass line around reactor recirculation system flow control valve was eliminated. Reduce the possibility of cavitation and cracking of piping in the recirculation system. Need eliminated by addition of low frequency motor generator set.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-18 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed Nuclear fuel The number of fuel pins in each fuel bundle has been changed from 7 x 7 to 8 x 8 (including two water rods). Improved fuel performance by increasing safety margins.

4.2 Nuclear

boiler A turbine building high temperature trip for MSIVs was added. Improve leak detection capability.

7.3.1 Nuclear

boiler An additional test mode was added for closing MSIVs one at a time to 90% of full open in the fast mode (close in slow mode already existed). Verifies that the spring force on the valves will cause them to close under loss-of-air

conditions.

5.4.5 Main steam line isolation A main condenser low vacuum initiation of the main steam line isolation was added. NRC requirement.

7.3.1 Main steam line isolation Reactor isolation was deleted for reactor high water

level. To provide improved plant availability.

5.4.5 Main steam line drain system A main steam line drain system was improved. Prevent accumulation of condensate in an idle line outboard of MSIV.

5.1.1 RPV code The RPV code was updated to ASME 1971 and Summer 1971 addenda. Update to applicable code as much as

possible.

5.2.1 Level

instrumentation The RPV level instrumentation was revised to eliminate Yarway columns and replace them with a conventional condensing chamber type; also, separation and redundancy features were added. Improve ECCS separation per IEEE-279

and improve reliability.

7.3.1 Turbine

seal setpoint

pressure The turbine seal setpoint pressure was changed from 50

psia to 125 psia. Ensures that main turbine condenser can extract reactor steam at temperature above

cooling capability of RWCU system. - Leak detection system The leak detection system was revised to upgrade the capability and incorporate the requirements of

IEEE-279. To meet IEEE-279 requirements.

7.6.1 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000LDCN-99-000 1.3-19 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed Reactor vibration

monitoring A confirmatory vibration monitoring test was added. NRC requirement.

14.2.12.3.34 RWCU system sample

station The P&IDs were changed to provide continuous

monitoring. Technical Specifications requirements. - LPCS system Valve F011 was changed from air-operated to motor-operated control. To provide Seismic Category I rated control

power to this essential active component. 7.3.1.1.1.3 LPCS system Direct connection to condensate storage replaced by removable spool piece connection to RHR. Condensate used only for system commissioning tests.

Figure 6.3-5 PRT replaced by RPT Prompt relief trip (PRT) was replaced by recirculation pump trip (RPT) for quick insertion of negative

reactivity. Increased reliability.

a 7.6.1.5 Main steam system Relief valve augmented bypass (REVAB) was deleted. Licensing requirement.

a - Feedwater sparger The thermal sleeve was changed to provide welded design of sparger to nozzle.

To eliminate vibration, failure, and leakage.

5.3 Standby

liquid control (SLC) system Interlocks on the SLC system were revised. To prevent inadvertent boron injection during system testing. 7.4.1, 9.3.5 Standby liquid control (SLC) system RCPB extended to explosive valves To meet isolation criteria. - RClC steam supply A warmup bypass line and valve was added. Permits pressurizing and prewarming of the steam supply line downstream to the turbine during reactor vessel heatup.

5.4.6 RCIC vacuum breaker system A vacuum breaker system was added to the RCIC

turbine exhaust line into the suppression pool. To prevent backup of water in the pipe and consequential high dynamic pipe loads and

reactions.

5.4.6 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-20 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed RCIC system Each component has been made capable of functional testing during normal plant operation. Improved testability.

5.4.6 Automatic

depressurization

system (ADS) The interlocks on the automatic depressurization system were revised. To meet lEEE-279 requirements.

7.3.1 RPV support The support for the RPV was changed from a ring girder to a bearing plate. Provides a better seismic and alignment

capability. 5.3.3.1.4.1 Plant service water pumps Upon loss of offsite power without a LOCA, the normal 4160 V service buses (SM-75, SM-85), are connected to SM-7 and SM-8 to provide automatic starting of a plant service water pump for drywell cooling. Provides service water for drywell cooling automatically after loss of offsite power without a LOCA.

Figure 8.1-2 , Tables 8.3-1 and 8.3-2 Reactor building cooling system ESF cooling units have been added to critical electric equipment areas in the reactor building. To provide suitable ambient temperature conditions for essential electrical and control equipment located in the reactor

building in the event of a LOCA.

9.4.9 Standby

gas treatment system Added second fan (powered from alternate power bus) to each standby gas treatment system. To remove need for crosstie between the two systems. 6.5.1.2 Standby gas treatment system Added facility to recirculate air from SGTS back into reactor building.

So that potential decay heat in filter can be removed without discharge to atmosphere in event of divisional power failure.

6.5.1.2 Standby gas treatment system Added second electric preheater (powered from

alternate power bus) to each SGTS unit. To provide means of controlling relative

humidity of air entering charcoal filter in event of primary heater or divisional power

failure. 6.5.1.2 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-21 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed Control room HVAC system Added two remote air intakes for pressurizing control room in event of a LOCA.

To limit doses to operating personnel to

limits of 10 CFR 50.

9.4.1.2 Ventilation system for areas in which essential cable is

routed Added to ESF ventilation system to ventilate corridors and cable chases through which essential cable is routed (diesel generators to control room). To provide suitable ambient temperatures

for essential cable in the event of a LOCA

9.4.8 Offgas

system charcoal vault Added a refrigeration system to the vault in which the offgas system charcoal adsorber filters are housed. To maintain charcoal adsorbers at a temperature of 0F. 9.4.5 , 11.3.2.1 Makeup water pumps transformer vault

ventilation Added a ventilation system to makeup water pump transformer rooms powered from the emergency buses. To ensure suitable ambient temperatures for transformers in the event of a loss of offsite power caused by a tornado.

9.4.6 Radioactive

waste

solidification process Cement-sodium silicate solidification process to be used in lieu of urea-formaldehyde process. To eliminate the generation of free water in solidified containers, a problem inherent to the urea-formaldehyde process.

11.4 Air ejector Three-stage air ejector to two-stage air ejector. Manufacturer offered a two-stage unit that meets the same operating conditions.

10.4.2 Sealing steam supply The gland steam evaporator will produce sealing steam using main steam on its tube side during startup and shutdown modes. PSAR stated auxiliary boiler would

be used. Adequate sealing steam can be produced with main steam pressure down to 125 psig.

10.4.3 Containment

instrument air The CIA air compressors were removed and the system is now supplied with nitrogen during reactor operation.

Redundant bottled gas supply utilized for supplying ADS valve accumulators for accident conditions. The purpose of the safety related bottled gas supplies is to back up the non-safety-related cryogenic nitrogen supply.

9.3.1.1.2 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.3-22 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed Offgas holdup line Radiography of circumferential welds was not done. A partial section of the line was buried before radiography was done. Welds were magnetic particle tested and line was hydro-tested at 1200 psig and then helium pressure decay leak tested with a sensitivity of 10-2 cm 3/sec. - Wet solid radwastes Packaged in 50 ft 3 containers rather than 50-gal drums. Reduce handling time and operator exposure.

11.4.2.10 Turbine bypass valve system Four bypass valves are used rather than three. Solution to operating problems with bypass valves on Cooper Nuclear station.

10.4.4 Main steam isolation valve leakage control system Added to plant. NRC requirement.

6.7 Main steam line from outermost isolation valve to turbine stop

valve Piping has been upgraded from Code Group D to Code

Group B. NRC requirement.

10.3.2 Radwaste tank sizes l. Waste sludge phase separator From 12,500 to 13,000 gal. To increase capacity.

Table 11.4-4 2. Chemical waste tank From 13,000 to 15,000 gal. To increase capacity. Table 11.2-13 3. Decontamination solution concen-trated waste tank From single 700-gal to two 700-gal tanks. To provide spare tank.

Table 11.4-4 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000LDCN-99-000 1.3-23 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed

4. Concentrated waste measuring

tank From 100 to 400 gal.

Due to increase in shipment container size from 50 gal to 50 ft

3. Table 11.4-4 5. Condensate phase separators From 12,500 to 23,500 gal. To increase capacity in event of higher than normal backwash requirements.

Table 11.4-4 6. Chemical addition tank From single 1000-gal tank to two 200-gal tanks. To provide capability for both acid and caustic addition from separate tanks.

Original tank oversized. Table 11.2-13 Floor drain system Influent waste radionuclide concentration changed from range of10

-4 to 10-2 Ci/ml to on order of 10

-1 Ci/ml. Reevaluation of source terms.

11.2.2.2.2 Liquid radwaste system GALE code was used to calculate radioactive discharges with 2500-gpm blowdown. Blowdown of 4000 gpm was used in the PSAR. NRC requirement to use GALE Code.

Change in blowdown results in more conservative (higher) radionuclide concentrations.

11.2.3.2 Cleaning of filters Changed from steam cleaning connections to chemical cleaning system.

Design improvement.

Figure 10.4-5 Missiles from

tornadoes Selection of credible missiles. For FSAR, followed specific missiles identified in NRC Standard Review Plan.

3.5.1.4 Cleaning of filters Changed from steam cleaning connections to chemical cleaning system.

Design improvement.

Figure 10.4-5 Missiles from

tornadoes Selection of credible missiles. For FSAR, followed specific missiles identified in NRC Standard Review Plan.

3.5.1.4 Primary containment

vessel New loads due to hydro-dynamic effects of safety/relief valve actuation and LOCA (neither in PSAR or FSAR; see Dynamic Analysis Report). To accommodate new GE load

requirements.

3.8.2 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000LDCN-99-000 1.3-24 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed Diesel generator

building fire protection system Changed from CO 2 system to dry pipe preaction system.

after a fire.

To provide accessibility to the diesel immediately. Also availability of unlimited water supply Appendix F Cable chase fire protection system Added dry pipe preaction system for cable chase and diesel generator building corridor. To protect divisional cable concentrations in

these areas.

Appendix F 500-kV line Hookstick changed to motor-operated switch. Available standard switches are supplied with motor operators. Fig. 8.1-2 500-kV line Line terminates at H. J. Ashe Swtichyard rather than Hanford Switching Station. BPA revisions to 500 kV grid.

8.1.2 230-kV line Deleted hookstick and 230-kV OCB at plant switchyard. OCB relocated to H. J. Ashe Switchyard. Fig. 8.1-2 115-kV line Replace circuit interrupter with 115-kV OCB at plant switchyard. Equipment availability. Fig. 8.1-2 Backup source Utilized to supply essential loads during diesel generator testing. PSAR did not consider particulars of diesel generator testing. 8.3.1.1.7.1.7 Diesel generator

starting Deleted automatic starting due to startup or backup transformer undervoltage. Class 1E bus undervoltage is the only

undervoltage condition requiring diesel generator start 8.3.1.1.7.1.7 8.3.1.1.7.2.7 Diesel generator trips during emergency

operation Added incomplete sequence trip to Division 1 and 2 diesel generators. Incomplete sequence indicates a diesel generator malfunction having an imminent possibility of unit damage. 8.3.1.1.7.1.8 125-V, 250-V-dc battery capability Revised supply capability from 4 hr to 2 hr. Increased dc load 8.3.2 125-V, 250-V-dc charger capability Revised recharge capability from 8 hr to 24 hr. Increased dc load 8.3.2 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011LDCN-11-008 1.3-25 Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continued)

Item Change Reason for Change FSAR Portion in Which Change is Discussed Spare 125-V-dc

charger Spare charger serves as a backup for Divisions 1 and 2

only. Spare charger is too large to provide

backup to Division 3. 8.3.2 Communication systems The commercial telephone exchange system is not redundant. Redundancy not required. 8.2.1.5 Fuel pool cooling and cleanup system Upgraded cooling portion of system to Seismic Category I to provide long-term cooling and safety grade makeup water capability for coolant of spent fuel

following refueling. To prevent fuel pool boiling and resultant adverse environmental conditions which could affect safety-related electrical equipment in the reactor building. 9.1.3 a PRT and REVAB were proposed at the CP stage as non-safety-related power generation type systems to reduce the thermal-hydraulic effects of transient events in the core. However, during experiments in the MK-11 suppression pool dynamics test program, it was decided that less frequent relief valve cycling during plant operation was desirable. Consequently, the recirculation pump trip (RPT) system was developed to perform functions previously assigned to PRT and REVAB.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDCN-03-026 1.4-1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS

The italicized information is historical and was provided to support the application for an

operating license.

1.4.1 APPLICANT/OPERATOR

Energy Northwest is a municipal corporation and a joint operating agency of the State of Washington, organized in January 1957, pursuant to Chapter 43.52 of the Revised Code of Washington, as amended. Energy Northwest assumes the responsibility for safe operation and maintenance of the plant and for providing related services as described in Chapter 13. 1.4.2 ENGINEER AND CONSTRUCTION MANAGEMENT - BURNS & ROE, INC.

Burns and Roe, Inc. (B&R) provides engineering and initial construction management and quality assurance services for the design and construction of the plant, integrating the major plant items furnished by the General Electric Company (GE) and Westinghouse Electric

Corporation.

Burns & Roe was founded in 1932 and incorporated in 1935 as Burns and Roe, Inc. Burns &

Roe has been active in the fields of power generation and distribution, sea water and brackish water desalination, waste water renovation, and engineering, design, and/or construction

management services for over 50 thermal power generating units representing more than

11,400,000 kW of new generating capacity, of which more than 4,800,000 kW is nuclear.

Burns & Roe, Inc., has been continuously engaged in construction of engineering activities

since 1935.

1.4.3 NUCLEAR

STEAM SYSTEM SUPPLIER - GENERAL ELECTRIC COMPANY

General Electric designed, fabricated, and delivered the direct-cycle boiling water nuclear steam supply system (NSSS) for Columbia Generating Station (CGS). General Electric also fabricated the first core of nuclear fuel and provided technical direction of installation and

startup of this equipment.

General Electric has engaged in the development, design, construction, and operation of boiling water reactors (BWRs) since 1955. Table 1.4-1 lists GE reactors completed, under construction, or ordered (several later canceled). Thus, GE has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the

installation and startup of reactors.

General Electric continues to provide technical support for the operation of CGS as requested by Energy Northwest. This includes providing support for the CGS Megawatt Improvement

Program (see Section 1.1).

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 1.4-2 1.4.4 TURBINE GENERATOR SUPPLIER - WESTINGHOUSE ELECTRIC CORP.

Westinghouse Electric Corporation designed, fabricated, delivered, and installed the turbine generator for CGS. They also provided technical assistance for the startup of this equipment.

Westinghouse Electric Corporation has a long history in the application of turbine generators in nuclear power stations going back to the inception of commercial electrical power

production using nuclear facilities. Westinghouse furnished the turbine generator unit for Shippingport No. 1. This unit was shipped in 1956. Westinghouse also furnished the turbine generator unit for Yankee Atomic Power Company Rowe No. 1. This unit was shipped in 1959. San Onofre No. 1 and Connecticut Yankee, Haddam Neck No. 1 unit went into commercial operation in 1968. Westinghouse nuclear turbine generators produced over 300 billion kW hr of electricity through May 1976, when 25 nuclear turbine generators totaling over 16,500 MW were in service. By 1984, 75 Westinghouse nuclear turbine generators

should be in service producing over 61,319 MW. Inlet steam pressures of these units vary between 750 psig and 1000 psig and electrical outputs vary from 500,000 kW to 1,090,000 kW.

Westinghouse continues to provide technical and maintenance support for the turbine generator on an as-requested basis. They also provided replacement for the three low-pressure turbine

rotors installed in the Spring 1992 refueling outage.

1.4.5 SYSTEM

COMPLETION CONTRACTOR - BECHTEL

As System Completion Contractor, Bechtel provides field and home office services in project planning and control, engineering, construction completion, startup support, and quality

verification for CGS. The Bechtel organization was founded in 1898, in the midwest, by

Warren A. Bechtel. In 1940, Bechtel went international, working on a pipeline system in Venezuela; and then vastly diversified its activities during World War II, becoming involved in naval bases, shipyards, pipelines, refineries, and aircraft modification. Next, Bechtel pioneered in the nuclear power field, constructing the first reactor to produce useful electricity in 1949, and building Dresden I, the first commercial nuclear power station. Today, Bechtel is recognized as one of the worlds leading engineering and construction firms.

1.4.6 MAJOR

CONTRACTORS 1.4.6.1 Fischbach/Lord

Fischbach/Lord is responsible for the major electrical installation at CGS, consisting of

raceways, conduit, cable, terminators, and electrical equipment. They were formed as a joint venture, solely for this project, in 1974.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-15-011 1.4-3 1.4.6.2 Pittsburgh-Des Moines Steel Company

Pittsburgh-Des Moines Steel Company is responsible for engineering, fabrication, and

installation of materials in the Primary Containment Vessel.

1.4.6.3 Wright-Schuchart- Harbor/Boecon (Boeing Construction)/General Energy Resources, Inc.

Wright-Schuchart-Harbor/Boecon/General Energy Resources, Inc. (WBG) was formed as a joint venture October 1, 1977, to be responsible for installation of major mechanical equipment, power, and process piping for CGS.

1.4.6.4 Bechtel

During plant construction, Bechtel served as the Construction Manager. During the operating phase Bechtel, as the Site Support Services contractor, is providing field engineering and installation support for plant modifications. Also, as Technical Services contractor, they are providing engineering support under Energy Northwest direction and under the Energy Northwest quality assurance program as requested by Energy Northwest. Under these contracts Bechtel is providing support to the Megawatt Improvement Program (see

Section 1.1).

1.4.6.5 Deleted

1.4.6.6 Westinghouse Electric

Westinghouse provided the turbine generator. They provided replacement of the three low-pressure rotors which were installed in 1992. Westinghouse also provided a new plant

simulator which was installed in 1995.

1.4.7 CONSULTING

ENGINEER - R. W. BECK AND ASSOCIATES

The independent consulting firm of R. W. Beck and Associates is the consulting engineer for Energy Northwests Columbia Generating Station. This firm was also a consulting engineer

for WNP-1. Having extensive experience in preparing engineering feasibility and financing studies and reports necessary for the success of utility and civic improvement projects, the firm is well qualified for employment as a consulting engineer and was chosen as a result of its

experience.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December2003 1.4-4 The duties of the consulting engineer are briefly summarized as follows: prepare estimates of plant capability, energy potential, usability within area loads and resources, the cost of power and energy output of the project, and generally determine the feasibility of the project. These duties will include assisting in preparation of a Bond Resolution, preparation of an engineering

report, schedules for investment of funds, schedules for debt service payments, and other engineering services necessary to facilitate the financing of the project.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.4-5 Table 1.4-1 Commercial Nuclear Reactors Completed, Under Construction, or in Design by General Electric Station Utility Rating (MWe) Year of Order Year of Startup Dresden 1 a Commonwealth Edison 207 1955 1960 Humboldt Bay a Pacific G&E 63 1958 1963 Kah1 a Germany 15 1958 1961 Garigliano a Italy 150 1959 1964 Big Rock Point Consumers Power 71 1959 1965 JPDR Japan 11 1960 1963 KRB a Germany 237 1962 1967 Tarapur 1 India 190 1962 1969 Tarapur 2 India 190 1962 1969 GKN Holland 52 1963 1968 Oyster Creek JCP&L 620 1963 1969 Nine Mile Point 1 Niagara Mohawk 610 1963 1969 Dresden 2 Commonwealth Edison 794 1965 1970 Pilgrim 1 Boston Edison 655 1965 1972 Millstone 1 NUSCo 660 1965 1970 Tsuruga Japan 340 1965 1970 Nuclenor Spain 440 1965 1971 Fukushima 1 Japan 439 1966 1971 BKW KKM Switzerland 306 1966 1972 Dresden 3 Commonwealth Edison 794 1966 1971 Monticello Northern States 536 1966 1971 Quad Cities 1 Commonwealth Edison 789 1966 1972 Browns Ferry 1 TVA 1065 1966 1974 Browns Ferry 2 TVA 1065 1966 1975 Quad Cities 2 Commonwealth Edison 789 1966 1972 Vermont Yankee Vermont Yankee 514 1966 1972 Peach Bottom 2 Philadelphia Electric 1065 1966 1974 Peach Bottom 3 Philadelphia Electric 1065 1966 1974 James A. FitzPatrick New York Power Authority 821 1966 1975 Bailly b NIPSCo 660 1966 ---- Shoreham b LILCo 819 1967 1985 Cooper Nebraska PPD 778 1967 1974 Brown Ferry 3 TVA 1065 1967 1977 Limerick 1 Philadelphia Electric 1055 1969 1985 Hatch 1 Georgia 786 1967 1975 Fukashima 2 Japan 762 1967 1974 Brunswick 1 Carolina P&L 790 1968 1977 Brunswick 2 Carolina P&L 790 1968 1975 Arnold Iowa ELP 545 1968 1975 Fermi 2 Detroit Edison 1056 1968 1984 Limerick 2 Philadelphia Electric 1055 1969 ----

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-99-000 1.4-6 Table 1.4-1 Commercial Nuclear Reactors Completed, Under Construction, or in Design by General Electric (Continued)

Station Utility Rating (MWe) Year of Order Year of Startup Hope Creek 1 PSE&G 1067 1969 1986 Hope Creek 2 b PSE&G 1067 1969 ----

Zimmer b CCDPP 810 1969 ---- Chinshan Taiwan 610 1969 1977 Caorso 1 Italy 827 1969 1975 Hatch 2 Georgia 795 1970 1979 LaSalle County 1 Commonwealth Edison 1078 1970 1983 LaSalle County 2 Commonwealth Edison 1078 1970 1984 Susquehanna 1 Pennsylvania P&L 1050 1968 1983 Susquehanna 2 Pennsylvania P&L 1050 1968 1984 Chinshan 2 Taiwan 610 1970 1978 Columbia Generating Station Energy Northwest 1103 1971 1984 Nine Mile Point 2 Niagara Mohawk 1090 1971 1986 Grand Gulf 1 Midsouth 1250 1972 1985 Kaiseraugst b Switzerland 915 1971

---

Fukushima Japan 1135 1971 1976 Takai 2 Japan 1135 1971 1976 River Bend 1 Gulf States 940 1971 1985 River Bend 2 b Gulf States 940 1971 ----

Perry 1 Cleveland Electric 1205 1971 1985 Perry 2 b Cleveland Electric 1205 1971 ---- Hartsville A-1 b TVA 1233 1972 ----

Hartsville B-1 b TVA 1233 1972 ----

Hartsville A-2 b TVA 1233 1972 ----

Hartsville B-2 b TVA 1233 1972 ---- Laguna Verde 1 Mexico 660 1972 1977 Leibstadt Switzerland 940 1972 1978 Kuosheng 1 Taiwan 992 1972 1978 Kuosheng 2 Taiwan 992 1972 1979 Clinton 1 Illinois Power 950 1973 1986 Clinton 2 b Illinois Power 950 1973

---

Montague 1 b NUSCO 1150 1973

---

Allens Creek 1 b Houston L&P 1200 1973 ----

Skagit 1 b Puget SD 1288 1973 ---- Skagit 2 b Puget SD 1288 1973 ---- Barton 3 b Alabama 1220 1973 ---- Blackfox 1 b Oklahoma 1150 1973

---

Blackfox 2 b Oklahoma 1150 1973

---

Cofrentes Spain 975 1973 1977 Laguna Verde 2 Mexico 660 1973 1978 Enel 6 b Italy 982 1974 ---- Enel 8 b Italy 982 1974 ---- a Retired b Discontinued C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-99-000 1.5-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION

The italicized information is historical and was provided to support the application for an

operating license.

1.5.1 GENERIC

ISSUES

NUREG-0933, A Prioritization of Generic Safety Issues presents the generic issues as

follows: a. TMI action plan items In NUREG-0933, these follow the content and format of NUREG-0660 and

NUREG-0737.

b. Task action plans

These include both the unresolved safety issues (USIs) previously included in NUREG-0606 and the Category A Generic Activities previously included in

NUREG-0371 and the Category B, C and D Generic Activities previously

included in NUREG-0471.

c. Human factors

These are the human factors considerations of NUREG-0660 and NUREG-0737.

d. Chernobyl Issues

This part addresses the recommendations of NUREG-1251.

In the sections below, these issues are addressed as unresolved safety issues (USIs), generic safety issues (GSIs), and TMI Task Action Plans. Human Factors considerations are included as part of the TMI Task Action Plans. Chernobyl is not addressed below or on the CGS docket as NUREG-1251 lead to the conclusion that no immediate changes in NRC regulations regarding the design or operation of U.S. commercial reactors were required. However, NUREG-1251 and INPO SER 34-86, Chernobyl Unit 4 Accident, and INPO SOER 87-1, Core Damaging Accident Following an Improperly Conducted Test, were reviewed by Energy Northwest to identify the need for any changes to hardware, procedures, or training at CGS.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-99-000 1.5-2 1.5.1.1 Unresolved Safety Issues

1.5.1.1.1 Unresolved Safety Issues Introduction

Unresolved safety issues are issues identified by the NRC that affect a number of plants, question the adequacy of existing requirements, have no current resolution and are judged to

be unacceptable if left unresolved for the life of the plant.

A December 20, 1977, amendment to the Energy Reorganization Act required that the NRC develop a plan providing for specification and analysis of USIs and take action as necessary to implement corrective measures with respect to such issues. In a joint Explanatory Statement of the House - Senate Conference Committee for the FY 1978 Appropriations Bill this was explained to mean that a plan was to be developed to resolve the USIs. In September 1989, the NRC achieved resolutions of all of the identified USIs.

On October 19, 1989, the NRC issued Generic Letter 89-21, Request for Information Concerning Status of Implementation of Unresolved Safety Issue (USI) Requirements. This generic letter requested that licensees and construction permit holders review and report on the status of the implementation of USIs for which final technical resolution had been achieved.

Energy Northwest responded to this request in Reference 1.5-1. The NRC responded to this submittal by Reference 1.5-2 and identified anticipated transient without scram (ATWS), Station Blackout and Safety Implications of Control Systems (A-9, A-44, and A-47, respectively) as not being implemented. (Subsequently, these have been resolved as discussed

below.)

1.5.1.1.2 Implementation of Specific Unresolved Safety Issues

A-8 Mark II Containment Pool Dynamic Loads

Resolution of A-8 for CGS is documented in NUREG-0892 (the SER for CGS) and Supplements 4 and 5 in Sections 6.2.1.8 and 3.9.3.1, respectively.

A-9 Anticipated Transients Without Scram In the safety evaluation transmitted with Reference 1.5-7 , the NRC stated that the standby liquid control (SLC) flow and sodium pentaborate decahydrate concentration for CGS were in compliance with the ATWS rule.

The design requirements for resolution of ATWS for CGS were to install an alternate rod

injection (ARI) system (see Section 7.4.1.6), a standby liquid control (SLC) system (see Sections 7.4.1.2 and 9.3.5), and to trip the reactor recirculation pumps automatically by a recirculation pump trip (RPT) system under conditions indicative of an ATWS (Section C OLUMBIA G ENERATING S TATION Amendment 56 F INAL S AFETY A NALYSIS R EPORT December 2001 LDCN-01-00A 1.5-3 7.4.1.5). In addition, ATWS equipment needed to be qualified for the environmental conditions associated with anticipated operational occurrences and to ATWS conditions up to the time the required function is completed (Reference 1.5-10). The FSAR Section 15.8 ATWS analysis also needed to be revised.

In Reference 1.5-3, the NRC stated that the CGS alternate rod injection system was in compliance with the ATWS rule. The reference also stated that the RPT system required two

modifications to be in compliance with the rule. Reference 1.5-4 documents the implementation of the changes required to resolve these two issues.

In Reference 1.5-5, Energy Northwest informed the NRC that confirmation of the environmental qualifications of ATWS equipment remained to be confirmed. Reference 1.5-6 documented that the confirmation had been completed.

In FSAR Amendment 42, Section 15.8 was revised to include new ATWS analyses. Technical Specification Amendment 93 was issued on August 9, 1991 which addressed modifications to the ATWS-RPT system. With this amendment, all activities required for ATWS resolution for CGS were completed.

A-10 BWR Feedwater Nozzle Cracking

NRC review of CGS relative to A-10 and NUREG-0619, which Generic Letter 89-21 states resolves this USI, is documented in NUREG-0892, Sections 3.9.3.1, 5.2.3.1, and 5.2.4. While

these sections address A-10, they do not specifically state that the total issue is resolved for

CGS. However, as no concerns were raised in the subsequent five supplements to

NUREG-0892 and as Energy Northwest was not aware of a concern of the NRCs regarding A-10 subsequent to the issuance of the operating license, in Reference 1.5-1 Energy Northwest stated that it believed A-10 to be resolved for CGS. This position was apparently accepted by

the NRC by the issuance of Reference 1.5-2.

A-11 Reactor Vessel Material Toughness

NRC acceptance of the CGS commitment to 10 CFR 50, Appendix G, is discussed in

NUREG-0892, Section 5.3.2. In NUREG-0744 and Generic Letter 82-26 issued subsequent to the publication of the original issue of NUREG-0892, a response by licensees was not required; they only provided guidance to licensees who may have been required to submit a fracture analysis to justify continued operation. This was not the case for CGS.

A-17 Systems Interactions

Generic Letter 89-18 issued September 6, 1989 transmitted NRC final resolution of this USI.

No formal reply was required. Energy Northwest incorporated information contained and

referenced in this Generic Letter into the CGS IPE program, the results of which were C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 1.5-4 submitted to the NRC by Reference 1.5-22. However, as no formal action to Generic Letter 89-18 was required, Energy Northwest considered this USI closed for CGS prior to the completion of the IPE. This was so stated in Reference 1.5-1.

A-24 Qualification of Class 1E Safety Related Equipment

In NUREG-0892 Supplement 4, Section 3.11.5, the NRC states that CGS has demonstrated conformance to NUREG-0588. Generic Letter 89-21 states that Revision 1 to NUREG-0588

resolved A-24. By NRC memorandum, J. Knight to T. Novak, dated November 1983 (8312120370), Mr. Knight states that the CGS review was to Revision 1 of the NUREG.

A-31 Residual Heat Removal Shutdown Requirements

NUREG-0892 states in Section 5.4.2.1 that the CGS RHR system conforms to the Commissions regulations and applicable Regulatory Guides. Generic Letter 89-21 states that

A-31 was resolved in May 1978 by publication of SRP 5.4.7. As NUREG-0892 was written in

May 1982, Energy Northwest stated in Reference 1.5-1 that this established closure of A-31 for CGS.

A-36 Control of Heavy Loads

NUREG-0892 Supplement 4, Section 9.1.5, states that the guidelines of NUREG-0612 have been satisfied for CGS. Generic Letter 89-21 states that NUREG-0612 resolves A-36.

A-39 Determination of Safety Relief Valve Pool Dynamic Load and Temperature Limits

Section 6.2.1.8 of NUREG-0892 Supplements 1 and 4, provides NRC acceptance of the resolution of this issue for CGS.

A-40 Seismic Design Criteria

NUREG-1233 issued September 1989 states that the proposed changes that constitute the

resolution of USI-40 are to apply to new applicants only. CGS is not one of the plants identified in Generic Letter 89-21 that needed to be reviewed to the new criteria.

A-42 Pipe Cracks in Boiling Water Reactors NUREG-0892 states in Section 5.2.3.1 that CGS conforms to the requirements of

NUREG-0313, Revision 1, which Generic Letter 89-21 states resolves A-42. NUREG-0892 Supplement 5, Section 5.2.3.2, provides additional information on this issue. Also see Section

5.2.3.2.3. Additional consideration for BWR pipe cracks beyond the scope of A-42 were raised by the NRC in Generic Letter 88-01. The resolution of Generic Letter 88-01 for CGS C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-99-000 1.5-5 is provided in References 1.5-21 , 1.5-35 , and 1.5-36 , and in the Bases for CGS Technical Specifications.

A-43 Containment Emergency Sump Performance

Generic Letter 89-21 states that resolution of A-43 only applies to new plants (i.e., those reviewed after October 1985) and, as such, does not apply to CGS.

A-44 Station Blackout

See Appendix 8A.

A-45 Shutdown Decay Heat Removal

According to guidance provided in Generic Letter 89-21 and Supplement 9 to NUREG-0933, Energy Northwest incorporated closure of A-45 into the CGS IPE program the results of which were submitted to the NRC by Reference 1.5-22.

A-46 Seismic Qualification of Equipment in Operating Plants

Generic Letter 87-03 issued February 27, 1987 which addresses A-46 resolution for CGS did

not require any action or plant review. NUREG-1211, Enclosure 1, established Generic Letter

87-03 as applicable to CGS rather than Generic Letter 87-02. As such, Energy Northwest considers this USI closed for CGS. Also, NUREG-0892, Supplement 5 in Appendix C states that A-46 only applies to plants that were operating at the time.

A-47 Safety Implication of Control System

Generic Letter 89-19 provides requirements to close A-47. The overfill protection system

required of BWRs is provided for in CGS. Closure of this issue was provided by Reference 1.5-9. A-48 Hydrogen Control Measures and Effects of Hydrogen Burn on Safety Equipment

As stated in Generic Letter 89-21, A-48 is closed and implemented for Mark II BWRs such as

CGS.

1.5.1.1.3 Unresolved Safety Issues Implementation Summary

The resolution of all USIs for CGS has been achieved with the NRC. Regarding Station

Blackout (A-44), 10 CFR 50.63(c)(4) provides for a 2 year implementation schedule for closure of identified modifications.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-99-000 1.5-6 1.5.1.2 Generic Safety Issues

1.5.1.2.1 Generic Safety Issues Introduction

In Generic Letter 90-04, Reference 1.5-12, the NRC requested that licensees and construction permit holders address a list of specific generic safety issues (GSIs) listed in the generic letter.

Energy Northwests response to this request for CGS was provided in Reference 1.5-13.

1.5.1.2.2 Implementation of Specific Generic Safety Issues The following summarizes the CGS implementation of applicable GSIs listed in Generic Letter

90-04 and other GSIs that have been resolved for CGS subsequent to the issuance of the Generic Letter. The following is a summary of information provided in Reference 1.5-13 with updated information provided as appropriate.

GSI/Subject Status 40/BWR Scram System Pipe Breaks Closed as documented in NUREG-0892 (p. 4-4) and documents listed in

Reference 1.5-13 41/BWR Scram Discharge Volume Closed as documented in NUREG-0892 (p. 7-6) 43/Reliability of Air Systems Closed as discussed in References 1.5-13 and 1.5-15 48/LCOs for Class 1E vital Instrumentation Buses - Generic Letter 91-11 (added

subsequent to Generic Letter 90-04

response)

Closed as documented in Reference 1.5-19 49/Interlocks and LCOs for Class 1E Tie Breakers - Generic Letter 91-11 (added

subsequent to Generic Letter 90-04

response).

Closed as documented in Reference 1.5-19 51/Improved Reliability of Open-Cycle

Service Water Systems Closed subsequent to Generic Letter 90-04

as addressed by References 1.5-11 , 1.5-37 , and 1.5-38 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.5-7 67/Improved Accident Safety Report Monitoring Closed as summarized in NRC Evaluation for CGS Regulatory Guide 1.97 implementation (Reference 1.5-14) 75/Salem ATWS Events Closed subsequent to the Generic Letter 90-04 response by letters listed in Reference 1.5-13 , Reference 1.5-17 , and issuance of Technical Specification Amendment 90.

Generic Letter 83-28, Supplement 1, issued

October 7, 1992, did not change this status as CGS does not use reactor trip breakers. 79/RPV Thermal Stress During Natural Convection Cooldown Closed subsequent to Generic Letter 90-04 by Generic Letter 92-02 as not impacting BWRs 86/Long Range Plan for Stress Corrosion

Cracking in BWR Piping Closed based upon documents listed in

Reference 1.5-13. A-13/Snubber Operability Assurance NUREG-0933 states that this issue was resolved in 1980 by revision to the Standard Technical Specifications (STS). As the

original CGS Technical Specifications

were based upon Revision 3 to the BWR STS

issued in 1980, this concern is resolved for CGS. In particular, for the five issues

mentioned for GSI A-13 resolution in Generic Letter 90-04: 1. The arbitrary capacity limit of 50,000 lbs that previously existed in Technical Specifications does not appear in the CGS Technical Specifications. 2. The requirement for NRC approval of seal material does not appear in the CGS Technical Specifications. 3, 4. Monitoring and IST programs to ensure snubber reliability do exist in the CGS Licensee Controlled Specifications. They are significantly expanded from that included in earlier programs.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.5-8 5. The CGS Licensee Controlled Specifications allow for an in-place snubber IST program.

Thus, the five requirements of A-13 resolution as discussed in Generic Letter 90-04 have been implemented for CGS A/30 Adequacy of Safety Related DC Power Supplies - Generic Letter 91-06 (added

subsequent to Generic Letter 90-04

response)

Closed as documented in Reference 1.5-18 A-35/Adequacy of Offsite NUREG-0892

Power Systems Closed as documented in NUREG-0892 (p. 8-16) and discussed in Reference 1.5-13) B-63/Installation of Low Pressure Systems

Connected to the RCPB Closed as discussed in Question 040.079 (FSAR Volume 22) and Reference 1.5-13 1.5.1.2.3 Generic Safety Issues Implementation Summary

Implementation of the applicable GSIs of Generic Letter 90-04 is complete.

1.5.1.3 TMI Task Action Plans

The CGS responses to the TMI-2 action plans as they were included in NUREG-0737 are provided in Appendix B. This Appendix agrees with Reference 1.5-16 in documenting that all TMI Task Action Plans have been implemented for CGS.

1.

5.2 REFERENCES

1.5-1 Letter, GO2-89-215, G. C. Sorensen to NRC, Response to Generic Letter 89-21 Requesting Plant Status on Implementation of Unresolved Safety Issues,

dated November 30, 1989.

1.5-2 Letter, R. B. Samsworth (NRC) to G. C. Sorensen (SS), Unimplemented Unresolved Safety Issues at WNP-2 (TAC No. 74538), dated March 20, 1990.

1.5-3 Letter, R. B. Samworth (NRC) to G. C. Sorensen (SS), ATWS Rule 10 CFR 50.62 relating to ARI and RPT Systems, dated November 6, 1988.

1.5-4 Letter, GO2-90-110, G. C. Sorensen to NRC, Anticipated Transients Without Scram (ATWS) Design Modifications, dated June 22, 1990.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.5-9 1.5-5 Letter, GO2-89-110, G. C. Sorensen (SS) to NRC, Anticipated Transients Without Scram Implementation Schedule, dated June 16, 1989.

1.5-6 Letter, GO2-90-116, G. C. Sorensen (SS) to NRC, Resolution of ATWS for WNP-2, dated June 29, 1990.

1.5-7 Letter, R. B. Samworth (NRC) to G. C. Sorensen (SS), Issuance of Amendment No. 43, dated May 29, 1987.

1.5-8 Letter, GO2-89-062, G. C. Sorensen (SS) to NRC, Response to Station Blackout Rule using HPCS Diversion III as Alternate AC Power, dated

April 17, 1989.

1.5-9 PL Eng (NRC) to G. C. Sorensen (SS), Response to Request for Action Related to Resolution of Unresolved Safety Issue A Safety Implications of Control System in LWR Nuclear Power Plants, pursuant to 10 CFR 50.54(f) - Generic Letter 89-19 (TAC NO. 75019), dated November 13, 1991.

1.5-10 BWROG Topical Report NEDE-31096-P, Anticipated Transients Without Scram; Response to NRC ATWS Rule 10 CFR 50.62, dated December 1985.

1.5-11 Letter, PL Eng (NRC) to G. C. Sorensen (SS), Evaluation of Response to NRC Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment (TAC No. 74086), dated April 26, 1992.

1.5-12 Generic Letter 90-04, Request for Information on the Status of Licensee Implementation of Generic Safety Issues Resolved With Imposition of Requirements or Corrective Actions, dated April 25, 1990.

1.5-13 Letter, GO2-90-113, G. C. Sorensen to NRC, Response to Generic Letter 90-04 Regarding Status of Implementation of Generic Safety Issues, (TAC No.

75993), dated June 28, 1990.

1.5-14 Letter, G. W. Knighton (NRC) to G. C. Sorensen (SS), Emergency Response Capability - Conformance to Regulatory Guide 1.97, Revision 2, (TAC No.

59516), dated March 23, 1988.

1.5-15 Letter, GO2-89-128, G. C. Sorensen to NRC, Final Response to Generic Letter 88-14, Instrument Air Supply Problems Affecting Safety-Related Equipment,

dated July 28, 1989.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-99-000 1.5-10 1.5-16 NUREG-1435 Supplement 2, Status of Safety Issues at Licensed Power Plants, dated December 1992.

1.5-17 Letter, P. L. Eng (NRC) to G. C. Sorensen (SS), Response to Generic Letter 90-03 for Washington Nuclear Plant 2 (TAC No. 76314), dated

November 8, 1990.

1.5-18 Letter, W. M. Dean (NRC) to G. C. Sorensen (SS), Response to Generic Letter 91-06, MPA L106, Resolution of Generic Issue A-30, Adequacy of Safety Related DC Power Supplies, Pursuant to 10 CFR 50.54(f) for Washington Public Power Supply System Unit 2 (TAC NO. M81515), dated

March 27, 1992.

1.5-19 Letter, P. L. Eng (NRC) to G. C. Sorensen (SS), Response to Generic Letter 91-11, Resolution of Generic Issues 48-LCOs for Class 1E Vital Instruments Buses and 49 - Interlocks and LCOs for Class 1E Tie Breakers pursuant to 10 CFR 50.54(f) for Washington Public Power Supply System Nuclear Plant No. 2 (TAC No. M82484), dated March 2, 1992.

1.5-20 Letter, P. L. Eng (NRC) to G. C. Sorensen (SS), Status of TMI Item I.D.1.2, Detailed Control Room Design Review (DCRDR) at Washington Public Power Supply System Nuclear Project No. 2 (WNP-2) (TAC No. 56181), dated

November 13, 1991.

1.5-21 Letter, P. L. Eng (NRC) to G. C. Sorensen (SS), Response to GL 88-01, Intergranular Stress Corrosion in Piping (TAC No. 69161), dated

December 28, 1990.

1.5-22 Letter, GO2-92-206, G. C. Sorensen (SS), Response to Generic Letter 88-20, Individual Plant Examinations for Severe Accident Vulnerabilities 10 CFR

50.54(f), dated August 28, 1992.

1.5-23 through 1.5-34 Deleted

1.5-35 Letter, GO2-92-004, G. C. Sorensen to NRC, Response to NRC SER on Generic Letter 88-01 (TAC No. 69161), dated January 8, 1992.

1.5-36 Letter, GO2-92-086, G. C. Sorensen to NRC, Additional Response to Generic Letter 88-01 Safety Evaluation Report (TAC Nos. M80358 and M69161), dated

April 10, 1992.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.5-11 1.5-37 Letter, GO2-90-017, G. C. Sorensen to NRC, Response to Generic Letter 89-13, Service Water System Problem Affecting Safety-Related Equipment, dated February 5, 1990.

1.5-38 Letter, GO2-91-041, G. C. Sorensen to NRC, Response to Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment,

dated February 28, 1991.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCE Table 1.6-1 is a list of General Electric topical reports and other reports and documents which are incorporated in whole or in part by reference. These documents were filed with the NRC.

C OLUMBIA G ENERATING S TATION Amendment53 F INAL S AFETY A NALYSIS R EPORT November1998

Table 1.6-1 Topical Reports Report Title FSAR Section 1.6-3 General Electric Company APED-4824 Maximum Two-Phase Vessel Blowdown from Pipes (April 1965) 6.2 APED-5458 Effectiveness of Core Standby Cooling Systems for General Electric Boiling Water

Reactors (March 1968) 5.4 APED-5460 Design and Performance of General Electric BWR Jet Pumps (July 1968) 3.9 APED-5555 Impact Testing on Collet Assembly for Control Rod Drive Mechanism 7RDB144A (November 1967) 4.6 APED-5640 Xenon Considerations in Design of Large Boiling Water Reactors (June 1968) 4.1 APED-5652 Stability and Dynamic Performance of the General Electric Boiling Water Reactor (April

1969) 4.1 APED-5696 Tornado Protection for the Spent Fuel Storage Pool (November 1968) 3.3 , 3.5 , 9.1 APED-5706 Incore Neutron Monitoring System for General Electric Boiling Water Reactors (November 1968; revised April 1969) 7.6 APED-5750 Design and Performance of General Electric Boiling Water Reactor Main Steam Line

Isolation Valves (March 1969) 3.9 , 5.4 GEAP-5620 Failure Behavior in ASTM A106B Pipes Containing Axial Through-Wall Flaws (April 1968) 5.2 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-10-004,15-011 1.6-4 GEAP-10546 Theory Report for Creep-Plast Computer Program (January 1972) 4.1 GEAP-13197 Emergency Cooling in BWRs Under Simulated Loss-of-Coolant (BWR PLECMP)

Final Report (June 1971) 6.2 GE-NE-778-028-0790 GE Duralife 215 Control Rod Safety Evaluation, Revision 2 (July 1992) 4.2 GE-NE-187-24-0992 Washington Public Power Supply System Nuclear Project 2, SRV Setpoint Tolerance

and Out-of-Service Analysis, Revision 2 (July 1993) 6.3 NEDC-31984-P Generic Evaluations of General Electric Boiling Water Reactor Power Uprate -

(July 1991) 5.4 , 15.8 NEDC-32115-P Washington Public Power Supply System Nuclear Project 2, SAFER/GESTR-LOCA

Loss-of-Coolant Accident Analysis (September 1992) 6.3 NEDC-32141-P Power Uprate With Extended Load Line Limit Safety Analysis for WNP-2 (June 1993) 5.4 , 15.8 NEDC-32232-P WNP-2 Reactor Recirculation Adjustable Speed Drive (ASD) System Reliability Analysis (August 1993) 7.7 NEDC-32983-P-A General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux

Evaluations (January 2006) 4.3.2.8 , 4.3.4 NEDC-33507-P Energy Northwest Columbia Generating

Station APRM/RBM/Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA), Revision 1 (January 2012) 15.4 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-15-011 1.6-5 NEDE-10169 Safe-System Analysis for Standby Core Cooling Equipment (September 1970) 3A NEDE-10313-P PDA - Pipe Dynamic Analysis Program for Pipe Rupture Movement 3.6 NEDE-11146-P Design Basis for New Gas System (July 1971) 11.3 NEDE-13442-P-01 Mark II - Pressure Suppression Test Program (May 1976) 3A NEDE-20944-P BWR/4 and BWR/5 Fuel Design (October 1976)

Table 1.3-1 NEDE-21175-3-P Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and

Loss-of-Coolant Accident (LOCA) Loadings (July 1982) 3.9 NEDE-21354-P BWR Fuel Channel Mechanical Design and Deflection (September 1976) 3.9 NEDE-21471-P Analytical Model for Estimating Drag Forces on Rigid Submerged Structures Caused by

LOCA and Safety/Relief Valve Ramshead Air Discharges (September 1977) 3A NEDE-21544-P Mark II Pressure Suppression Containment System, an Analytical Model of the Pool

Swell Phenomenon (December 1976) 3A, 6.2 NEDE-21821 BWR Feedwater Nozzle/Sparger Final Report (March 1978) 5.2, 5.3 NEDE-23604 Brunswick Unit 1 Reactor Internals Vibration and Temperature Measurements (June 1977) 5.3 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-13-013 1.6-6 NEDE-23749-P Analytical Model for Computing Transient Pressure and Forces in the S/RVDL (February 1978) 3.9 NEDE-23806-P MK II Main Vent Lateral Loads Summary Report (October 1978) 3A NEDE-24010-P Technical Bases for the Use of the SRSS Method for Combining Dynamic Loads for Mark II Plants (July 1977) with Supplement 1 (October 1978), Supplement 2 (December 1978), and Supplement 3 (August 1979) 3.9 NEDE-24011-P-A General Electric Standard Application for Reactor Fuel (most recent approved version referenced in COLR) 1.8 , 3.9 , 4.1 , 4.2 , 4.3 , 4.4 , 15.1 , 15.4 NEDE-24057-P Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants (November 1977) 3.9 NEDE-24106-P Dynamic Lateral Loads on a Main Vent Downcomer - Mark II Containment (March 1978) 3A NEDE-24222 Assessment of Boiling Water Reactor Mitigation of Anticipated Transient Without

Scram, Volume II (December 1979) 15.8 NEDE-24285-P Chugging Loads - Revised. Definition and Application Methodology for Mark II

Containments (July 1981) 3A NEDE-24288-P Generic Condensation Oscillation Load Definition Report (November 1980) 3A NEDE-24302-P Generic Chugging Load Definition Report (April 1981) 3A C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORT December2007 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-06-000 1.6-7 NEDE-24695 RVF0R04 Users Manual, S/RVDL Clearing Transient Pressures and Forces in the S/RDL (December 1979) 3.9 NEDE-24794-P Dynamic Lateral Loads on Mark II Main Vent Downcomer - Correlation of Independent Reference Data (March 1980) 3A NEDE-24811-P 4T Condensation Oscillation Test Program Final Test Report (May 1980) 3A NEDE-24822-P Mark II Improved Chugging Methodology (May 1980) 3A NEDE-24834 Hanford 2 Crimped Control Rod Drive Line (June 1980) 3.6 NEDE-24988-P Analysis of Generic BWR Safety/Relief Valve Operability Test Results (October 1981) 5.2 , 5.4 , Table F.4-1 NEDE-25100-P CAORSO SRV Discharge Tests Phase I Test Report (May 1979) 3A NEDE-25118 CAORSO SRV Discharge Tests Phase II ATR Report (August 1979) 3A NEDE-31096-P Licensing Topical Report, Anticipated Transient Without Scram (February 1987) 4.6 , 7.4 , 9.3 NEDM-10320 The General Electric Pressure Suppression Containment Analytical Model (March 1971) 3A , 6.2 NEDO-10029 An Analytical Study on Brittle Fracture of GE BWR Vessel Subject to the Design Basis

Accident (July 1969) 1.8 NEDO-10320 The General Electric Pressure Suppression Containment Analytical Model (April 1971) 3A 6.2 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-13-013 1.6-8 NEDO-10329 Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors (April 1971);

Supplement 1, (April 1971); Addenda, (May 1971) 6.2 NEDO-10349 Analysis of Anticipated Transients Without Scram (March 1971) 15.8 NEDO-10466-A Power Generation Control Complex Design Criteria and Safety Evaluation (September 1977) 8.3 , 9.5 , F.2 , F.3 , F.7 NEDO-10527 Rod Drop Accident Analysis for Large Boiling Water Reactors (March 1972);

Supplement 1, (July 1972); Supplement 2, (January 1973) 4.2 , 15.4 NEDO-10602 Testing of Improved Jet Pump for the BWR/6 Nuclear System (June 1972) 3.9 NEDO-10734 A General Justification for Classification of Effluent Treatment System Equipment as

Group D (February 1973) 11.3 NEDO-10751 Experimental and Operational Confirmation of Off-Gas System Design Parameters (January 1973) (Proprietary) 11.3 NEDO-10802 Analytical Methods of Plant Transient Evaluations for General Electric Boiling Water Reactor (February 1973) 15.2 NEDO-10899 Chloride Control in BWR Coolants (June 1973) 1.8 , 5.2 NEDO-10905 HPCS Power Supply 1.8 , 8.3 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-15-011 1.6-9 NEDO-10951 Releases from BWR Radwaste Management Systems (July 1973) 11.2 NEDO-10958-A General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application (January 1977) 15.0 NEDO-20533 The General Electric Mark III Pressure Suppression Containment System Analytical

Model (June 1974) 3A , 6.2 NEDO-20566-A Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K (Proprietary) (September 1986) 3.9 , 4.2 , 6.3 NEDO-20626 Studies of BWR Designs for Mitigation of Anticipated Transients without Scrams (October 1974) 6.2 , 9.3 NEDO-20761 Millstone Nuclear Power Station, Refueling/Maintenance Outage (Fall 1974) 12.2 NEDO-21061 Mark II Containment Dynamics Forcing Functions Information Report (September 1976, June 1978, November

1981) 3A , 6.2 NEDO-21142 Realistic Accident Analysis for General Electric Boiling Water Reactor - The RELAC Code and Users Guide (December 1977) 15.2 , 15.6 NEDO-21231 Banked Position Withdrawal Sequence (September 1976) 15.4 NEDO-21471 Analytical Model for Estimating Drag Forces on Rigid Submerged Structures Caused by

LOCA and Safety/Relief Valve Ramshead Air Discharges (September 1977) 3A C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-10-029 1.6-10 NEDO 21667 Comparison of the 1/13 Scale Mark II Containment Multivent Pool Swell Data with Analytical Methods (August 1977) 3A NEDO-21708 Radiation Effects in Boiling Water Reactor Vessel Steels (October 1977) 5.3 NEDO-21778-A Transient Pressure Rises Affecting Fracture Toughness Requirements for Boiling Water Reactors January 24, 1978 (January 17, 1979) 5.3 NEDO-21985 Functional Capability Criteria for Essential Mark II Piping (September 1978) 3.9 NEDO-23678-P Mark II Pressure Suppression Test Program Phases I, II, and III of the 4T Tests (June

1978) 3A NEDO-24057-P Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants (November 1977) 3.9 NEDO-24154-A Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, Volumes 1 and 2 (August 1986) 5.2 NEDO-24210 PISYS Analysis of NRC Problem (August 1979) 3.9 NEDO-24226 General Electric Company, Control Blade Lifetime With Potential B 4 C Loss, with Supplement 1 (December 1979) 4.2 NEDO-24288 Mark II Containment Program - Generic Condensation Oscillation Load Definition Report (February 1981) 3A NEDO-24548 Technical Description Annulus Pressurization Load Adequacy Evaluation (January 1979) 6.2 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-15-011 1.6-11 NEDO-24708-A Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors (June 1980) 7.4 , B , I , Table F.4-1 NEDC-24154-P-A Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, Volumes 1, 2, 3 and 4 (February 2000) 15.0 , 15.1 , 15.2 , 15.3 , 15.5 NEDC-32084P-A TASC-03A A Computer Program for Transient Analysis of a Single Channel (July 2002) 6.3 NEDC-32601P-A Methodology and Uncertainties for Safety Limit MCPR Evaluations (August 1999) 4.4 NEDC-32694P-A Power Distribution Uncertainties for Safety Limit MCPR Evaluations (August 1999) 4.4 NEDC-32868P GE14 Compliance With Amendment 22 of NEDE-24011-P-A (GESTAR II) (May 2013) 4.1 , 4.2 , 4.3 , 4.4 , 15.4 NEDE-32906P Supplement 3-A Migration to TRACG04/PANAC11 from TRACG02/PANAC10 for TRACG AOO and ATWS Overpressure Transients (April 2010) 5.2 NEDC-32950P Compilation of Improvements to GENEs SAFER ECCSLOCA Evaluation Model (July 2007) 6.3 NEDC-33270P GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR 11) (May 2013) 4.1 , 4.2 , 4.3 , 4.4 , 9.1 , 15.4 NEDE-23785-1-PA The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant

Accident. Volumes 1, 2, and 3 (October 1984) 6.3 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-15-011 1.6-12 NEDE-23785P-A The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident. Volume 3 Supplement 1, Additional Information for Upper Bound PCT

Calculation. (March 2002) 6.3 NEDE-24011-P-A-US General Electric Standard Application for

Reactor Fuel (GESTAR II) (Supplement for United States) (most recent approved version

referenced in COLR) 3.9 , 4.1 , 4.2 , 4.3 , 4.4 , 15.4 NEDE-30130-P-A Steady State Nuclear Methods (April 1985) 15.1 , 15.4 C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section LDCN-15-062 1.6-13 Other References WPPSS-74-2-R2 and Supplements WPPSS-74-2-R2A and

WPPSS-74-2-R2B Washington Public Power Supply System Sacrificial Shield Wall (March 1974)

Sacrificial Shield Wall Design Supplemental

Information (February 1975, August 1975) 3.8 , 6.2 Report Submitted

with letter

GO2-80-172, August 8, 1980 Engineering Evaluation of the WNP-2 Sacrificial Shield Wall (March 1974) 3.8 , 6.2 Report submitted with

letter GO2-80-182, August 19, 1980 Engineering Evaluation of the WNP-2

Sacrificial Shield Wall, Supplement No. 1 3.8 , 6.2 -- Plant Design Assessment Report for SRV and LOCA Loads 3A WPPSS-74-2-R3 Burns & Roe, Inc., Protection Against Pipe Breaks Outside Containment (April 1974) 3.5 WPPSS-74-2-R5 Drywell to Wetwell Leakage Study (July 1974, February 1974) (GO2-74-17, dated May 9, 1974) 6.2 , 3.8 Inservice Inspection

Program Plan Inservice Inspection Program Plan Interval - 4 5.2.4, 6.6 Preservice Inspection

Program Plan Preservice Inspection Program Plan 5.2.4, 6.6 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Table 1.6-1 Topical Reports (Continued)

Report Title FSAR Section 1.6-14 CGS-FTS-0168 Columbia Generating Station Alternative Source Term (report consolidated from letters

GO2-04-170 dated September 30, 2004, GO2-06-116 dated September 11, 2006, GO2-05-054 dated March 16, 2005, GO2-05-160 dated September 29, 2005, GO2-06-043 dated March 21, 2006, GO2-06-105 dated August 7, 2006 and GO2-06-108 dated August 24, 2006) 1.8 , 15.4 , 15.6 , 15.7 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 1.7-1 1.7 ACRONYMS

The acronyms used in this FSAR follow

ACI American Concrete Institute

ACRS Advisory Committee on Reactor Safeguards

ADS automatic depressurization system AEC Atomic Energy Commission

AISC American Institute of Steel Construction

ALARA as low as is reasonably achievable

ALI annual limit on intake

AMP Aging Management Programs ANSI American National Standards Institute

APRM average power range monitor

ARM area radiation monitor

ART adjusted reference temperature

AS auxiliary steam

ASCE American Society of Civil Engineers

ASD adjustable speed drive

ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials

ATWS anticipated transient without scram

AWS American Welding Society

B&R Burns and Roe, Inc.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 1.7-2 BISI bypass & inoperable status indication

BOC beginning of cycle

BPA Bonneville Power Administration

BPC Bechtel Power Corporation

BWR boiling water reactor BWROG Boiling Water Reactor Owners Group BWRVIP BWR Vessel and Internals Project

CAS central alarm station, control air system CASS Cast Austenitic Stainless Steel

CEP containment exhaust purge

CGS Columbia Generating Station

CHF critical heat flux

CIA containment instrument air CLB current licensing basis

CMFA common mode failure analysis

COLR Core Operating Limits Report

COND main condensate system

CPR critical power ratio CRA primary containment cooling system

CRD control rod drive

CRDA control rod drop accident CRDRL control rod drive return line C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035,12-020 1.7-3 CREF control room emergency filtration

CRPI control rod position indication

CSP containment purge supply

CST condensate storage and transfer, condensate storage tank CUF Cumulative Usage Factor

CW circulating water

DAC derived air concentrations

DAW dry active radioactive waste

DB design basis

DBA design basis accident

DBE design basis earthquake

DG diesel generator

DEH digital electrohydraulic DLR dosimeter of legal record

DOE Department of Energy

DOP dioctylphthalate DSA Diesel Starting Air

DZO depleted zinc oxide ECA engineering change authorization

ECCS emergency core cooling system

ECN engineering change notice

EDR equipment drain (radioactive) processing C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 1.7-4 EFCV excess flow check valve

EFPY Effective Full Power Years

EHC electrohydraulic control EMA Equivalent Margin Analysis

EOC end of cycle EOF emergency operations facility

EPA electrical protection assembly

EPN equipment piece number EPRI Electric Power Research Institute

EPZ emergency planning zone EQ Environmental Qualification, Environmentally Qualified

ESF engineered safety feature

EWD electrical wiring diagram

FA full arc (mode of TGV operation)

FAC Flow Accelerated Corrosion

FANP Framatome ANP

F-B/V front to back/vertical

FCD functional control diagram FCV flow control valve

FDDR Field Deviation Disposition Request

FDR floor drain (radioactive) processing system

FLECHT full-length emergency cooling heat transfer C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-15-033 1.7-5 FMEA failure modes effects analysis

FPC fuel pool cooling and cleanup

FSAR Final Safety Analysis Report

GE General Electric Company

HAD heat actuated device HCA horizontal control accelerometer

HCU hydraulic control unit

HCV Hardened Containment Vent

HELB high energy line break

HEPA high-efficiency particulate air/absolute

HID high-intensity discharge (lighting--vapor lamp)

HPCS high-pressure core spray

H&V heating and ventilating

HVAC heating, ventilating, and air conditioning

HX heat exchanger

IASCC Irradiation Assisted Stress Corrosion Cracking

IBA intermediate break accident

IDC incident detection circuitry IDS instrument data sheet

IED instrument engineering diagram

IEEE Institute of Electrical and Electronics Engineers

IGA Intergranular Attack C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.7-6 IGSCC intergranular stress corrosion cracking

IHSI Induction Heat Stress Improvement

IRM intermediate range monitor

ISA Instrument Society of America

ISI In-Service Inspection ISP Integrated Surveillance Program

LCO Limiting Condition of Operation

LCS leak control system

LDS leak detection system

LHGR linear heat generation rate

LLRT local leak rate test

LOCA loss-of-coolant accident

LPCI low-pressure coolant injection

LPCS low-pressure core spray

LPRM local power range monitor

LPZ low population zone

LRA License Renewal Application

LSSS limiting safety system setting MAPLHGR maximum average planar linear heat generation rate

MCC motor control center

MCPR minimum critical power ratio

MEL Master Equipment List C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.7-7 MG motor-generator

MLD mean low water datum

MLHGR maximum linear heat generation rate

MOV motor-operated valve

MS main steam MSIV main steam isolation valve

MSIV-LCS main steam isolation valve leakage control system

msl mean sea level

MSL main steam line

MSLC main steam isolation valve leakage control

MWR mixed waste (radioactive)

MWt Megawatt thermal

NB nuclear boiler

NBR nuclear boiler rated (power)

NDE nondestructive examination

NDT nil-ductility transition

NDTT nil-ductility transition temperature

NEC National Electrical Code NED Nuclear Energy Division (GE)

NFPA National Fire Protection Association

NEPIA Nuclear Energy Property Insurance Association

NMS neutron monitoring system C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 1.7-8 NPDES National Pollutant Discharge Elimination System

NPHS net positive suction head

NRC Nuclear Regulatory Commission

NSAS non-safety affecting safety

NSOA nuclear safety operational analysis NSSS nuclear steam supply system

NSSSS nuclear steam supply shutoff system

OBE operating basis earthquake

OQAPD Operational Quality Assurance Program Description

ODCM Offsite Dose Calculation Manual

OPRM Oscillation Power Range Monitor

OSHA Occupational Safety and Health Act

OT operating transient

OS&Y outside screw and yoke

OT operating transient

PA Public Address (System)

PABX Private Automatic Branch Exchange

PATP Power Ascension Test Program PCIOMR preconditioning cladding interim operating management recommendation

PCRVICS primary containment and reactor vessel isolation control system

PCS process computer system C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 1.7-9 PCT peak cladding temperature

PDIS plant display information system

PEC Plant Engineering Center

PGCC power generation control complex

P&ID piping and instrumentation diagram PMF probable maximum flood

PPM Plant Procedure Manual

PRM power range monitor

PSAR Preliminary Safety Analysis Report

PSF Plant Support Facility

PVS plant vent stack

RBM rod block monitor

RCC reactor building closed cooling water

RCIC reactor core isolation cooling

RCPB reactor coolant pressure boundary

REA reactor building exhaust air

RFW reactor feedwater

RHR residual heat removal RMC reactor manual control

RMS remote manual switches

ROA reactor building outside air C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-14-004 1.7-10 RPIS rod position information system

RPS reactor protection system

RPT recirculation pump trip

RPV reactor pressure vessel

RRC reactor recirculation system RRS required response spectra

RSO reactor system outline

RT NDT Reference Temperature for Nil-Ductility Transition RWCU reactor water cleanup

RWM rod worth minimizer

RWP Radiation Work Permit

SA service air

SACF single active component failure

SAF single active failure

SAR Safety Analysis Report

SAS Secondary Alarm Station

SBA small break accident

SBO station blackout SCC Stress Corrosion Cracking

SCF single component failure

SCC/IGA SCC/intergranular attack C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 1.7-11 SCC/IASCC SCC/irradiation assisted stress corrosion cracking

SDC shutdown cooling

SEF single equipment failure

SER Safety Evaluation Report

SF single failure (NSOA)

SGT standby gas treatment

SGTS standby gas treatment system

SJAE steam jet air ejector

SLC standby liquid control

SLMCPR minimum critical power ratio safety limit

SLO single loop operation

SMS Scheduled Maintenance System

SOE single operator error

SPC Siemens Power Corporation

SPV solenoid pilot valve

SRM source range monitor

SRO Senior Reactor Operator

SRP Standard Review Plan

SRV safety/relief valve

SS safe shutdown

SS stainless steel

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 1.7-12 SSC structures, systems, and components

SSE safe shutdown earthquake

S-S/V side-to-side/vertical

SSW sacrificial shield wall

SW standby service water SWP Site Wide Procedure

TCV turbine control valve

TDAS transient data acquisition system

TEDE total effective dose equivalent

TG turbine generator

TGV turbine governor valve

TIP traversing in-core probe TLAA Time Limited Aging Analysis

TLD thermoluminescent dosimeter

TMU tower makeup

TPM thermal power monitor

TRS test response spectra

TSC Technical Support Center TSPM Test and Startup Program Manual

TSW plant service water (turbine building service water)

TWG Test Working Group

UBC Uniform Building Code C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-09-035 1.7-13 UHS ultimate heat sink

UPS uninterruptable power supply

USE Upper Shelf Energy

WNP-2 Washington Nuclear Project No. 2

WPPSS Washington Public Power Supply System

ZPA zero period acceleration

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDCN-99-000 1.8-1 1.8 CONFORMANCE TO NRC REGULATORY GUIDES

1.

8.1 INTRODUCTION

This section of the FSAR contains information on Energy Northwests conformance assessment of CGS to Regulatory Guides, Division 1, Power Reactor Guides and revisions thereof as

noted.

Since the scope of equipment responsibility is project unique and the time of equipment design, procurement, manufacture, installation, and operation varies with the supplier, a unique assessment for the nuclear steam supply system (NSSS) scope of supply and balance of plant (BOP) scope of supply is necessary and is presented.

1.8.2 NUCLEAR

STEAM SUPPLY SYSTEM SCOPE OF SUPPLY EVALUATION

The following paragraphs define the nomenclature and the manner in which the NSSS scope of

supply assessment is to be interpreted.

Regulatory Guides - Incorporated in the Design

This section serves to identify specific safety or regulatory guides which were included in the plant as a design commitment during the construction permit review. It also identifies those incorporated by commitment after the construction permit issuance. All of these are specifically noted as Incorporated in the Design.

Regulatory Guides - Assessed Capability in the Design

For those other regulatory guides which have been issued before, during, or after the construction permit issuance, Energy Northwest (through his agents and/or suppliers) has performed an assessment evaluation to determine the capability of the previously approved design to accommodate and meet these new requirements. These are noted as Assessed Capability in the Design.

Conformance to the regulatory guide falls under either one of two categories - Full Compliance or Meeting Intent Through an Alternate Approach.

Regulatory Guide - Full Compliance Any regulatory guide so noted, whether by direct conformance or by assessed capability, complies with subject requirements as described in the FSAR.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009,06-014 1.8-2 Regulatory Guide - Meeting Intent by Alternate Approach

This designation is based on NRC rules which state that Regulatory Guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions different from those set out in the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permit or license by the Commission. The description and justification of an alternate approach is provided where this method is employed.

The following evaluation represents the NSSS scope of supply regulatory guide assessment.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.8-3 Regulatory Guide 1.1, Revision 0, November 1970 Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal

Pumps.

Regulatory Guide Intent:

This guide prohibits design reliance on pressure and/or temperature transients expected during a loss-of-coolant accident (LOCA) for ensuring adequate net positive suction head (NPSH). The requirements of this regulatory guide are applicable to the high-pressure core spray (HPCS), low-pressure core spray (LPCS), and residual heat

removal (RHR) pumps.

Applicable Assessment:

Incorporated in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in CGS is in

full compliance with this regulatory guide.

General Compliance or Alternate Approach Assessment:

The boiling water reactor (BWR) design conservatively assumes 0 psig containment pressure and maximum expected temperature of the pumped fluids; thus no reliance is placed on pressure and/or temperature transients to assure adequate NPSH.

Requirements for NPSH are available at the centerline of the pump suction nozzles for

each pump.

Specific Evaluation

Reference:

See Sections 6.2 and 6.3. Similar Application

Reference:

Similar application was used for LaSalle and GESSAR.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 1.8-4 Regulatory Guide 1.2, Revision 0, November 1970 Thermal Shock to Reactor Pressure Vessels

Regulatory Guide Intent:

This regulatory guide states that potential reactor pressure vessel brittle fracture which

may result from emergency core cooling systems (ECCS) operation need not be reviewed in individual cases if no significant changes in presently approved core and pressure vessel designs are proposed. Should it be concluded that the margin of safety against reactor pressure vessel brittle failure due to ECCS operation is unacceptable, and engineering solution, such as annealing, could be applied to ensure adequate recovery of the fracture toughness properties of the vessel material. This regulatory guide requires that available engineering solutions be outlined and requires that it be demonstrated that the design does not preclude their use.

Application Assessment:

Incorporated in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in CGS is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment:

The reactor pressure vessel used for CGS employs no significant core or vessel design changes from previously approved BWR pressure vessels such as Browns Ferry, all

units.

An investigation of the structural integrity of BWR pressure vessels during a design-basis accident (DBA) has been conducted (see NEDO-10029, An Analytical Study on

Brittle Fracture of GE-BWR Vessel Subject to the Design Basis Accident). It has been determined, based on methods of fracture mechanics, that no failure of the vessel by brittle fracture as a result of a DBA will occur.

The investigation included

a. A comprehensive thermal analysis considering the effect of blowdown and the LPCI system reflooding, C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 1.8-5 b. A stress analysis considering the effects of pressure, temperature, seismic load, jet load, dead weight, and residual stresses,

c. The radiation effect on material toughness [nil ductility transition temperature (NDTT) shift and critical stress intensity], and

d. Methods for calculating crack tip stress intensity associated with a nonuniform stress field following DBA.

This analysis incorporated very conservative assumptions in all areas (particularly in the areas of heat transfer, stress analysis effects of radiation on material toughness, and crack tip stress intensity). Therefore, the results reported in NEDO-10029 provide an upper bound limit on brittle fracture failure mode studies. Because of the upper bound approach, it is concluded that catastrophic failure of the pressure vessel due to the DBA

is shown to be impossible from a fracture mechanics point of view. In the case studies, even if an acute flaw does form on the vessel inner wall, it will not propagate as the

result of the DBA.

The criteria of 10 CFR 50, Appendix G, are interpreted as establishing the requirement for annealing. Paragraph IV C of Appendix G requires vessels to be designed for annealing of the beltline only where the predicted value of adjusted RT NDT exceeds 200°F as defined in paragraph NB2331 of the ASME Section III Code. This predicted value is not exceeded; therefore design for annealing is not required.

Specific Evaluation

Reference:

See Section 5.3.1.5. Similar Application

Reference:

Similar application was used for Browns Ferry 1, 2, and 3.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-04-050,05-055 1.8-6 Regulatory Guide 1.6, Revision 0, March 1971

Independence Between Redundant Standby (Onsite) Power Source and Between Their

Distribution Systems

Regulatory Guide Intent:

The guide states the extent and nature of independence of the two onsite power divisions required by General Design Criterion (GDC) 17. Key features that ensure operation and prevent cascading single failures from disrupting both power systems are delineated.

Application Assessment

Incorporated in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment

This regulatory guide is applicable to redundant standby (onsite) power sources and their distribution systems.

HPCS Onsite Power System (NSSS Scope of Supply)

Division 3 (HPCS) is provided with one offsite power source. Only one offsite supply

is connected because no credit is given to offsite power sources in accident analysis. The diesel generator breaker can be closed automatically only if the other source breakers to the (HPCS) load group are open.

When the HPCS diesel generator breaker is closed, no other source breaker can be closed automatically. No other means exist for automatically connecting redundant load groups with each other. The HPCS diesel generator may be manually connected to either Division 1 or to Division 2 in the extended station blackout (SBO) or non-DBA loss of offsite power (LOOP) scenario described in Section 8.3.1.1.7.2.1. The source breakers are administratively controlled in the open position to prevent paralleling of standby sources.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-055 1.8-7 Sufficient interlocks are provided to prevent paralleling the diesel generators manually by operator error during loss of offsite power. Division 3 diesel generator is provided

with only one prime mover.

The HPCS division dc load group is fed from one battery charger and one battery.

The HPCS standby power source and distribution system is independent from the other

two standby power sources and associated distribution system in the plant.

Specific Evaluation Reference

See Section 8.3.1.2. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-8 Regulatory Guide 1.9, Revision 0, March 1971 Selection of Diesel Generator Set Capacity for Standby Power Supplies

Regulatory Guide Intent:

This guide provides an approach for ensuring sufficient onsite power capability and for determining load requirements of diesel generator set power sources.

Application Assessment:

Incorporated in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment

This regulatory guide is applicable to the standby ac power supply for the HPCS diesel.

The specific guidelines are unduly restrictive when applied to the selection of the diesel generator set dedicated to the HPCS system. This is mainly due to the unique application of the special HPCS equipment relative to normal diesel generator units.

Specific conformance and alternate positions to and with Regulatory Guide 1.9 are

described in the following statements:

Regulatory Guide 1.9, Position 2 Conformance

Chapter 8 illustrates that the 2000-hr rating of the HPCS diesel generator, the 90% of 30-minute rating, and the maximum coincidental load, are in conformance with this position. Intermittent loads such as motor-operated valves are considered for long-term

loads. Regulatory Guide 1.9, Position 3 Conformance

CGS load requirements were verified as test data was completed and analyzed, following the preoperational tests.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-9 Regulatory Guide 1.9, Position 4 Conformance The HPCS diesel generator unit is considered as a unique application with justifiable departure from the strict conformance to Regulatory Guide 1.9, Revision 0, regarding voltage and frequency limits during the initial loading transient. The HPCS system consists of one large pump and motor combination which represents more than 90% of the total load; consequently, limiting the momentary voltage drop to 25% and the momentary frequency drop to 5% would not significantly enhance the reliability of HPCS operation. To meet the specific regulatory guide requirements, a diesel generator unit approximately two to three times as large as that required to carry the continuous rated load would be necessary. The specific diesel engine-electric generator-pump assembly was designed specifically for this integral operation. The frequency and voltage over-shoot requirements of Regulatory Guide 1.9, Revision 0, are met. A factory testing program on a prototype unit has verified the following

functions:

a. System fast-start capabilities,

b. Load-carrying capability,

c. Load shedding capability,

d. Ability of the system to accept and carry the required loads, and

e. The mechanical integrity of the diesel-engine generator unit and all of the major system auxiliaries.

GE Licensing Topical Report, HPCS Power Supply, NEDO-10905, describes the theoretical analytical aspects of the unique application including prototype and

reliability test considerations.

The design of the HPCS diesel generator conforms with the applicable sections of IEEE criteria for Class 1E Electrical Systems for Nuclear Power Generation Stations,

IEEE 308-1971.

The generator has the capability of providing power for starting the required loads with operationally acceptable voltage and frequency recovery characteristics. A partial or complete load rejection will not cause the diesel engine to trip on overspeed.

A special prototypic test conducted at the LaSalle facility verified the hardware and

load aspects of the HPCS power supply concept. This test is described in topical report

NEDO-10905, Revision 3.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-11-002 1.8-10 The scope of Regulatory Guide 1.9, Revision 0 does not include recommendations for surveillance testing. The surveillance requirements for demonstrating the operability of the diesel generators are consistent with the recommendations of Regulatory Guide 1.9 Revision 3 as described in the Bases for Technical Specification B 3.8.1. Compliance with Regulatory Guide 1.9 Rev. 0, as an acceptable basis for the selection of diesel generator sets of sufficient margin to implement General Design Criterion 17, remains as described herein.

Specific Evaluation

Reference:

See Section 8.3.1.2.1.4. Similar Application

Reference:

Similar application was used for LaSalle; for comparison see Table 8.3-6.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-11 Regulatory Guide 1.13, Revision 0, March 1971 Fuel Storage Facility Design Basis

Regulatory Guide Intent

This guide delineates design criteria that are appropriately applied to the fuel storage facility of the CGS plant.

Application Assessment

Incorporated in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General compliance or Alternate Approach Assessment

This regulatory guide is applicable to the refueling platform within NSSS scope of

supply.

The refueling platform is designed to prevent it from toppling into the pools during a safe shutdown earthquake (SSE). Redundant safety interlocks are provided as well as limit switches to prevent accidentally running the grapple into the pool walls.

Specific Evaluation Reference

See Section 9.1.4.3. Similar Application References

Similar application was used for Nine Mile Point 2.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-12 Regulatory Guide 1.20, Revision 2, May 1976 Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational

and Initial Startup Testing

Regulatory Guide Intent

Regulatory Guide 1.20 describes a comprehensive vibration assessment program for reactor internals during preoperational and initial startup testing. The vibration assessment program meets the requirements of Criterion 1, Quality Standards and Records, of Appendix A to 10 CFR Part 50 and Section 50.34, Contents of Applications; Technical Information, of 10 CFR Part 50.

Application Assessment

Incorporated in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General or Alternate Approach Statement

This regulatory guide is applicable to the core support structures and other reactor

internals.

A vibration measurement program has been defined for the confirmatory testing of this

plant during initial startup tests.

CGS reactor internals were tested in accordance with provisions of Regulatory Guide 1.20, Revision 2, Category IV, using Tokai-2 as the limited valid prototype.

Specific Evaluation Reference

See Sections 3.9.2.1 , 3.9.2.3 , and 3.9.2.4. Similar Application Reference

Similar application was used for Browns Ferry 1.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-13 Regulatory Guide 1.21, Revision 1, June 1974 Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear

Power Plants

Regulatory Guide Intent

Regulatory Guide 1.21 describes programs for measuring, reporting, and evaluating releases of radioactive materials in liquid and gaseous effluents and guidelines for classifying and reporting the categories and curie content of solid wastes.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment

The process and effluent radiological monitoring and sampling system is designed to provide the monitoring and sampling capability required to make the measurements, evaluations, and reports recommended by this guide.

The radiation monitoring systems (RMS) provided to meet these objectives are

a. For gaseous effluent streams Reactor building ventilation exhaust plenum RMS

b. For liquid effluent streams

1. Radwaste effluent RMS, and

2. Service water RMS

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-14 c. For gaseous process streams

1. Offgas pretreatment RMS,

2. Offgas posttreatment RMS, and

3. Carbon bed vault RMS

d. For liquid process streams

1. RHR service water RMS, and 2. Reactor building closed cooling water RMS These systems have the capability for alarm and initiation of automatic closure of waste treatment discharge valves in the affected systems prior to exceeding the normal operation limits specified in Technical Specifications thereby satisfying the intent of the

regulatory guide.

Specific Evaluation Reference

See Sections 7.6.1.1 and 11.5.1. Similar Application

Reference:

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-15 Regulatory Guide 1.22, Revision 0, February 1972 Periodic Testing of Protection System Actuation Function

Regulatory Guide Intent

This guide describes acceptable design approaches that facilitate the periodic testing, during reactor operation, of actuation devices/equipment incorporated into the reactor protection system design. This regulatory guide is applicable to the systems within NSSS scope of supply listed in this regulatory guide.

Application Assessment

Incorporated in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used for this facility is in full compliance with this regulatory guide.

General Compliance or Alternate Approach Assessment

Compliance for each system is discussed for this plant in the listed references.

Section

Reactor protection system 7.2.2.3 Emergency core cooling system HPCS 7.3.2.1.3 Automatic depressurization system (ADS) 7.3.2.1.3 LPCS 7.3.2.1.3 LPCI (RHR) 7.3.2.1.3 Primary containment and reactor vessel isolation 7.3.2.1.3 control system (PCRVICS)

Reactor core isolation cooling (RCIC) 7.4.2.3 Leak detection system 7.6.2.4 HPCS standby power supply 8.1.3 RHR system containment spray cooling system 7.3.2.1.3 Suppression pool cooling system 7.3.2.1.3 Reactor shutdown cooling system 7.4.2.3 Standby liquid control system 7.4.2.3 Process radiation monitoring system 7.6.2.4 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 1.8-16 Specific Evaluation Reference

See above.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-13-043,15-032 1.8-17 Regulatory Guide 1.26, Revision 3, February 1976

Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-

Containing Components of Nuclear Power Plants

Regulatory Guide Intent:

Regulatory Guide 1.26 describes a quality classification system for determining acceptable quality standards for important to safety components containing water, steam, or radioactive material other than those components addressed in 10 CFR

50.55a.

Application Assessment:

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

The definition of quality group classifications for this plant was made initially and recorded in the Preliminary Safety Analysis Report (PSAR) in accordance with ASME

Boiler and Pressure Vessel Code (B&PV), Sections III and VIII. Quality group classifications were maintained during design and construction and are actively

maintained during plant operations and modifications commensurate with the safety functions performed by the safety-related components.

This regulatory guide is applicable to Quality Groups B through D pressure parts including piping, pumps, valves, and vessels. Section

3.2 shows

the quality groups classifications of these parts.

Specific Evaluation

Reference:

See Section 3.2.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-000 1.8-18 Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-19 The italicized information is historical and was provided to support the application for an operating license.

Regulatory Guide 1.28, Revision 0, June 1972

Quality Assurance Program Requirements (Design and Construction).

Regulatory Guide Intent

This guide describes an acceptable method of complying with the NRCs regulations with regard to overall quality assurance program requirements.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The identified BWR Quality Assurance Program used in this facility reflects compliance with provisions of NRC regulations and regulatory guides or NRC-approved alternate

positions.

General Compliance or Alternate Approach Assessment

The General Electric BWR Quality Assurance Program has been developed over the years such that at any point in time it has been in compliance with mandatory regulatory requirements such as 10 CFR 50, Appendix B, and the ASME Code.

Implementation of the applicable ANSI-N45.2 series standards and the associated NRC regulatory guides (or NRC-approved GE alternate positions) has been an evolutionary

process and although partial implementation has always been effected before the date of issue of the regulatory guide or AEC Guidance on Quality Assurance, which recognized applicable ANSI standards, full implementation was not necessarily in place until the GE commitment date (see Attachment A for complete listing of GE commitment

dates and extent of commitment).

Since GE operates under a single quality assurance (QA) program, quality system improvements, such as more formalized audits or certification programs, are generally implemented across the board on all active projects with no opportunity for retrofit of completed work; therefore, work performed later in a project is typically subject to more quality assurance effort as a result of additional requirements. Attachment B gives a graphic representation of the time relation of some of the major project activities with the date of issue of regulatory guides and the GE commitment dates.

Because of the long generation cycle of the related ANSI Standard, GE had already C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-20 upgraded its QA program to at least partially implement each of the related ANSI Standards, where applicable, prior to the date of issue of the regulatory guide.

Attachment B also shows approximate dates of NRC and utility customer/architect-engineer QA audits. These audits have been performed frequently enough and over a

long enough time period to establish confidence that GE has been following a QA program which has kept current with customer and regulatory requirements.

Obviously, where most equipment is ordered years in advance of shipment, the QA program at the time of shipment will necessarily be somewhat different from that which was in effect at the time of ordering; however, at any point in time the GE QA program has been equal or better than the requirements in effect at that time.

Specific Evaluation Reference

Information was provided at the PSAR stage.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-014 1.8-21 Regulatory Guide 1.29, Revision 3, September 1978 Seismic Design Classification Regulatory Guide Intent

Regulatory Guide 1.29 describes an acceptable method of identifying and classifying those features of light-water-cooled nuclear power plants that should be designed to

withstand the effects of the SSE.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment:

This regulatory guide is used as a basis for defining the systems and components which must meet Seismic Category I requirements.

For the purpose of defining equipment that should be described to withstand the SSE, NSSS equipment conforms to the guide. The regulatory guide needs to be more specifically integrated in the following areas:

C.1(b) Application of this guide is limited to those reactor vessel internals which use engineered safety features, such as core spray piping, core spargers, and hardware, etc.

C.1(h) The component cooling water portions of the reactor recirculation pumps are not required to be Seismic Category I since the pumps do not perform a safety function.

Specific Evaluation Reference

See Section 3.2 , Table 3.2-1 , and the OQAPD.

Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-22 Regulatory Guide 1.30, Revision 0, August 1972 Quality Assurance Requirements for the Installation, Inspection, and Testing of

Instrumentation and Electric Equipment.

Regulatory Guide Intent

This guide describes an acceptable method of complying with the NRCs regulations

with regard to overall QA program requirements.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulatory guide or NRC regulations and

NRC-approved alternate positions.

General Compliance or Alternate Approach Assessment

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation Reference

Information was provided at the PSAR stage. Compliance is discussed in the OQAPD.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-23 Regulatory Guide 1.31, Revision 1, June 1973 Control of Stainless Steel Welding

Regulatory Guide Intent

Regulatory Guide 1.31 describes an acceptable method of implementing requirements with regard to the control of welding when fabricating and joining austenitic stainless

steel components and systems.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment:

All austenitic stainless steel weld filler materials were supplied with a minimum of 5%

delta ferrite. This amount of ferrite is considered adequate to prevent microfissuring in

austenitic stainless steel welds.

An extensive test program performed by GE, with the concurrence of the NRC, has demonstrated that controlling weld filler metal ferrite at 5% minimum produces

production welds which meet the requirements of this regulatory guide.

A total of approximately 400 production welds in five BWR plants were measured and all welds met the requirements of the Interim Regulatory Position.

Specific Evaluation Reference

See Section

5.2.3. Similar

Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-24 Regulatory Guide 1.32, Revision 1, March 1976 Use of IEEE 308-1974, Criteria for Class 1E Electric Systems for Nuclear Power Generating

Stations

Regulatory Guide Intent

This guide describes a method for implementation of electrical safety related equipment

design relative to GDC 17 and 18. This guide does contain some conflicts between GDC 17 and IEEE 308-1974 that of course will require resolution by plant design implementation. This regulatory guide is applicable to the battery and battery charger

of the HPCS standby power system.

Applicable Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

The HPCS battery charger has sufficient capacity to restore its battery to full charge under the maximum steady-state load within a 24-hr period. A period of 24 hr is

considered to be adequate to restore the battery from the design minimum charge state to the fully charged state irrespective of the status of the plant.

Specific Evaluation Reference

See Section 8.3.1.2. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-25 Regulatory Guide 1.34, Revision 0, December 1972 Control of Electroslag Weld Properties.

Regulatory Guide Intent

Regulatory Guide 1.34 describes an acceptable method of implementing requirements regarding control of weld properties when fabricating electroslag welds for nuclear components made of ferritic or austenitic materials.

Application Assessment

Not applicable.

Compliance or Alternative Approach Statement

Not applicable.

General Compliance or Alternate Approach Assessment

The electroslag welding process is not used on components within the NSSS scope of

supply. Therefore this regulatory guide is not applicable.

Specific Evaluation Reference

Not applicable.

Similar Application Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-26 Regulatory Guide 1.37, Revision 0, March 1973 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of

Water-Cooled Nuclear Power Plants.

Regulatory Guide Intent

This guide describes an acceptable method of complying with the NRCs regulations

with regard to overall QA program requirements.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulations and NRC regulatory guide or

NRC-approved alternate position.

General Compliance or Alternate Approach Assessment

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation Reference

Information was provided at the PSAR stage. Compliance is discussed in the OQAPD.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-27 Regulatory Guide 1.38, Revision 2, May 1977 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants.

Regulatory Guide Intent

This guide describes an acceptable method of complying with the NRCs requirements

for handling of nuclear materials.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulations and the NRC regulatory guide or

NRC-approved alternate position.

General Compliance or Alternate Approach Assessment

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation Reference

Information was provided at the PSAR stage.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-28 Regulatory Guide 1.41, Revision 0, March 1973 Preoperational Testing of Redundant On-Site Electric Power Systems to Verify Proper Load

Group Assignments.

Regulatory Guide Intent

The requirements of this regulatory guide are applicable to the total onsite electric power systems within Energy Northwests responsibility.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility

is in full compliance with this regulatory guide.

General Compliance or Alternate Approach Assessment

The HPCS power system is designed to be tested independently of any other redundant load group.

Specific Evaluation Reference

See Sections 8.3 and 14.2. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-29 Regulatory Guide 1.43, Revision 0, May 1973 Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components

Compliance or Alternate Approach Statement:

This regulatory guide is not applicable since stainless steel cladding on coarse grain

low-alloy steel for safety class components is not used.

General Compliance or Alternate Approach Assessment:

Not applicable.

Specific Evaluation

Reference:

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-30 Regulatory Guide 1.44, Revision 0, May 1973 Control of the Use of Sensitized Steel

Regulatory Guide Intent

The purpose of Regulatory Guide 1.44 is to address GDC 1 and 4 and 10 CFR 50 Appendix B requirements to control the application and processing of stainless steel to

avoid severe sensitization could lead to stress corrosion cracking. The guide proposes that this should be done by limiting sensitization due to welding as measured by ASTM A 262 Practice A or E, or another method that can be demonstrated to show

nonsensitization in austenitic stainless steels.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment

Tests by GE indicate that the test specified by A262 A or E (Detecting Susceptibility to Intergranular Attack in Stainless Steel) detects sensitization in a gross way, and the tests do not provide a precise method of predicting susceptibility to stress corrosion cracking in the BWR environment.

All austenitic stainless steel for CGS was purchased in the solution heat treated condition in accordance with applicable ASME and ASTM specifications. Carbon content was limited to 0.08% maximum, and cooling rates from solution heat treating

temperatures were required to be rapid enough to prevent sensitization.

Welding heat input was restricted to 110,000 joules per inch maximum, and interpass temperature was restricted to 305°F. High heat welding processes such as block welding and electroslag welding were not permitted. All weld filler metal and castings

were required by specification to have a minimum of 5% ferrite.

Whenever any wrought austenitic stainless steel was heated to temperatures over 800°F, by means other than welding or thermal cutting, the material was re-solution heat treated.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-31 These controls were used to avoid severe sensitization and to comply with the intent of Regulatory Guide 1.44.

Specific Evaluation Reference

See Section

5.2.3. Similar

Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-005 1.8-32 Regulatory Guide 1.45, Revision 0, May 1973

Reactor Coolant Pressure Boundary Leak Detection System.

Regulatory Guide Intent:

The guidelines are prescribed to ensure that leakage detection and collection systems provide maximum practical identification of leaks from within the reactor coolant pressure boundary (RCPB).

Application Assessment:

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment:

The leak detection system consists of temperature, pressure, fission product monitoring and flow sensors with associated instrumentation and alarms. This system detects, annunciates, and isolates (in certain cases) leakages in the following systems:

a. Main steam lines,

b. Coolant systems within the drywell,

c. Reactor water cleanup (RWCU) system,

d. RHR system,

e. RCIC system,

f. Feedwater system, and

g. HPCS system.

Leakage is separated into identified and unidentified categories thus meeting position C.1 of Regulatory Guide 1.45. The affected systems and the leakage detection methods are discussed in Section 5.2.5.1. Small unidentified leaks (5 gpm and less) inside the drywell are detected by temperature changes, pressure changes, drain sump pump activities, fission product monitoring, and floor drain flow monitoring; floor drain flow includes drywell cooler condensate flow.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-020 1.8-33 Large leaks are also detected by changes in reactor water level and changes in flow

rates in process lines.

The 5 gpm leakage rate is the limit on unidentified leakage inside the drywell. The leak detection system is capable of monitoring the flow rates with an accuracy of 1 gpm and is thus in compliance with paragraph C.2 of Regulatory Guide 1.45.

By monitoring drywell equipment and floor drain sump flow rates, which includes drywell coolers condensate flow rates and fission products (airborne particulate and gaseous radioactivity), position C.3 is satisfied.

Isolation and/or alarm of affected systems and the detection methods used are summarized in Table 5.2-12.

Monitoring of coolant for radiation in the Residual Heat Removal (RHR) and Reactor Water Cleanup (RWCU) heat exchangers satisfies position C.4 of the Regulatory

Guide. (For system details see Sections 7.6.1.2 and 11.5.)

The three methods differ in sensitivity and response time. Position C.5 requires the leak detection system be able to detect a leakage rate of 1 gpm in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. See

Section 7.6.2.4 for further discussion.

The leakage detection system instruments listed in Table 7.6-2 have been evaluated and are capable of performing their functions following an operating basis seismic event.

The drywell airborne particulate monitoring channel will remain functional following a safe shutdown earthquake. This satisfies position C.6 of Regulatory Guide 1.45.

Leakage detection indicators and alarms are provided in the main control room. This satisfies C.7 for the NSSS scope of supply. Procedures are developed for converting the various indications to a common leakage equivalent for the operators to satisfy

remainder of C.7.

The leakage detection systems are equipped with provisions to permit testing for operability and calibration during operation by the following methods:

a. Continuous monitoring of sump level compared to flow rates into sump,

b. Operability checked by comparing one method to another,

c. Simulation of signals into trip monitors, and

d. Channel A against Channel B of the same method.

Thus position C.8 is satisfied.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-005 1.8-34 Limiting conditions for identified and unidentified leakage are established as 20 gpm and 5 gpm respectively, thus satisfying position C.9.

Specific Evaluation

Reference:

See Sections 5.2.5 and 7.6.2.4. Similar Application

Reference:

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-35 Regulatory Guide 1.46, Revision 0, May 1973 Protection Against Pipe Whip Inside Containment

Regulatory Guide Intent:

Regulatory Guide 1.46 describes an acceptable basis for selecting the design locations

and orientations of postulated breaks in fluid system piping within the reactor

containment and for determining the measures that should be taken for restraint against pipe whipping that may result from such breaks.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment:

This regulatory guide is applicable to the recirculation pipe lines.

The design of the containment structure, component arrangement, Class 1 pipe runs, pipe whip restraints and compartmentalization was done in consonance with the acknowledgment of protection against dynamic effects associated with postulated rupture of piping. Analytically sized and positioned pipe whip restraints were

engineered to preclude damage based on the pipe break evaluation.

Pipe whip requirements for fluid system piping within the primary containment that, under normal operation, has service temperature greater than 200°F or pressures

greater than 275 psig, complied with ANS N176, Design Basis for Protection Against Pipe Whip, and Regulatory Guide 1.46 except as delineated in the following criteria for no breaks in Class 1 piping:

a. If Equation 10 of NB-365301, ASME Code Section III results in S<2.4 S m for ferritic or austenitic steels, no other requirements need be met. Stress range

should be calculated between any two load sets (including zero load set)

according to NB-3600 for upset and on operating basis earthquake (OBE) event

transient;

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-36 b. If Equation 10 results in 2.4<S<3.0 S m for ferritic or austenitic steels, the cumulative usage factor, U, calculated on the bases of Equation 14 of NB-3653.6, must be less than 0.1; and

c. If Equation 10 results in S>3.0 S m for ferritic or austenitic steels, then the stress value in Equations 12 and 13 of NB-3653.6 must not be greater than

2.4 S m. Specific Evaluation Reference

See Section

3.6. Similar

Application Reference

Similar application was used in GESSAR.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-15-008 1.8-37 Regulatory Guide 1.47, Revision 0, May 1973

Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems.

Regulatory Guide Intent:

This guide describes an acceptable method of complying with the requirements of IEEE 279-1971 and Appendix B to 10 CFR 50.

Application Assessment:

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of the regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment:

The alternate approach is provided in Section 7.1.2.4.

Specific Evaluation

Reference:

See Section 7.1.2.4. Similar Application

Reference:

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 1.8-38 Regulatory Guide 1.48, Revision 0, May 1973 Design Limits and Loading Combinations for Seismic Category I Fluid System Components.

Regulatory Guide Intent:

Regulatory Guide 1.48 provides acceptable design limits and appropriate combinations of loadings associated with normal operation, postulated accidents, and specified seismic events for the design of the Seismic Category I fluid system components.

Application Assessment:

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of

the alternate approach cited.

General Compliance or Alternate Approach Assessment:

For a comparison of NSSS with Regulatory Guide 1.48, see the attached tabulation.

The design basis was representative of good industry practices at the time of design, procurement, and manufacture and is shown to be in general agreement with requirements of Regulatory Guide 1.48, with the following clarifications:

a. The probability of an OBE of the magnitude postulated for CGS is consistent with its classification as an emergency event. However, for design conservatism, loads due to the OBE vibration motion have been included under upset conditions; loads due to the OBE vibratory motion plus associated transients, such as a turbine trip, have been considered in the equipment design under emergency conditions consistent with the probability of the OBE

occurrence; and

b. The use of increased stress levels for Class 2 components is consistent with industry practice as specified in ASME Code Section III.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 1.8-39 Specific Evaluation

Reference:

See Section

3.9.3. Similar

Application

Reference:

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007 1.8-41 COMPARISON WITH REGULATORY GUIDE 1.48 NRC Regulatory Guide 1.48 Columbia Generating Station

Component

Plant Condition

Loading Combination 1/

Design Limit Regulatory Guide Paragraph

Loading Combination (f)

Code allowable Stresses ASME Section III Reference How CGS Compares With NRC Regulatory Guide 1.48 Class 1 vessels Upset (U) (NPC or UPC) + 0.5 SSE NB-3223 1.a (NPC or UPC), 0.5 SSE 3.0Sm (includes NB-3223 Reflects industry secondary stresses) position Emergency (E) EPC NB-3224 2/ 1.b EPC, 0.5 SSE + transient 1.8Sm NB-3224 Faulted (F) NPC + SSE + DSL NB-3225 1.c NPC + SEE + DSL App.F-Sec. III NB-3225 Class 1 piping U (NPC or UPC) + 0.5 SSE NB-3654 1.a (NPC or UPC), 0.5 SSE 3.0Sm (includes NB-3654 Reflects industry secondary stresses) position E EPC NB-3655 2/ 1.b EPC, 0.5 SSE + transient 2.25Sm NB-3655 F NPC + SSE + DSL NB-3656 1.c NPC + SSE + DSL 3.0Sm NB-3656 Class 1 pumps U (NPC or UPC) + 0.5 SSE NB-3223 5/ 2.a (NPC or UPC), 0.5 SSE 1.65Sm NB-3223 Reflects industry (inactive) E EPC NB-3224 1/ 2.b EPC, 0.5 SSE + transient 1.8Sm NB-3224 position F NPC + SSE + DSL NB-3225 2.c NPC + SSE + DSL App. F-Sect. III NB-3225 Class 1 pumps U (NPC or UPC) + 0.5 SSE NB-3222 5/ 4.a.1 (NPC or UPC), 0.5 SSE Not Not Not (active) E EPC NB-3222 6/ 4.a.2 EPC applicable applicable applicable F NPC + SSE + DSL NB-3222 7/ 4.a.3 NPC + SSE + DSL 8/ Class 1 valves U (NPC or UPC) + 0.5 SSE NB-3223 5/ 2a (NPC or UPC), 0.5 SSE Not Not Not (inactive) by analysis E EPC NB-3224 4/ 2.b EPC applicable applicable applicable F NPC + SSE + DSL NB-3225 2/ 2.c NPC + SSE + DSL Class 1 valves U (NPC or UPC) + 0.5 SSE 1.1 Pr 3.a (NPC or UPC), 0.5 SSE 1.1 Pr NB-3525 Reflects industry (inactive) designed by E EPC 1.2 Pr 3.b EPC, 0.5 SSE + transient 1.2 Pr NB-3526 position either std. or

alternative F NPC + SSE + DSL 1.5 Pr 3.c NPC + SSE + DSL 1.5 Pr NB-3527 design rules Class 1 valves U (NPC or UPC) + 0.5 SSE NB-3222 5/ 4.a.1 (NPC or UPC, 0.5 SSE Not Not Not (active) by analysis E EPC NB-3222 6/ 4.a.2 EPC applicable applicable applicable F NPC + SSE + DSL NB-3222 7/ 4.a.3 NPC + SSE + DSL 8/ Class 1 valves (active) U (NPC or UPC) + 0.5 SSE 1.0 Pr 5.a.1 (NPC or UPC), 0.5 SSE 1.0 Pr NB-3525 Reflects industry designed by std. or E EPC 1.0 Pr 6/ 5.a.2 EPC 1.0 Pr (a) NB-3526 position alternative design rules F NPC + SSE + DSL 1.0 Pr 5.a.3 NPC + SSE + DSL 1.0 Pr NB-3527 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007 1.8-42 COMPARISON WITH REGULATORY GUIDE 1.48 (Continued)

NRC Regulatory Guide 1.48 Columbia Generating Station Component Plant Condition

Loading Combination 1/

Design Limit Regulatory Guide Paragraph

Loading Combination (f) Code Allowable Stresses ASME Section III Reference How CGS Compares With NRC Regulatory Guide 1.48 Class 2 & 3 vessels U (NPC or UPC) + 0.5 SSE1.1S 6.a (NPC or UPC), 0.5 SSE m = 1.1S code case 1607 Faulted condition, (Division 1) of section E EPC 1.1S 9/ 6.b EPC,0.5 SSE + transient (c)NC/NB NRC more conservative,VIII of the ASME Code F NPC + SSE + DSL 1.5S 6.c NPC + SSE + DSL m = 2.0S 3321.1(b) reflects industry position Class 2 vessels U (NPC or UPC) + 0.5 SSENB-3223 7.a (NPC or UPC), 0.5 SSE Not applicable Not applicable Not applicable (Division 2) of section E EPC NB-3224 2/ 7.b EPC VIII of the ASME Code F NPC + SEE + DSL NB-3225 7.c NPC + SSE + DSL Class 2 & 3 U (NPC or UPC) + 0.5 SSENC3611.1(b)(4)(c)(b)(1) 8.a (NPC or UPC), 0.5 SSE 1.2 Sh NC/ND 3611.3(b)NRC more conservative, piping E EPC NC3611.1(b)(4)(c)(b)(1) 10/ 8.a EPC,0.5 SSE + transient1.8 Sh NC/ND 3611.3(c)Reflects industry F NPC + SSE + DSL NC3611.1(b)(4)(c)(b)(2) 8.b NPC + SSE + DSL 2.4 Sh (4)(b) (b) position code case1606, NC/ND 3611.3(d)

[see note (b)] Class 2 & 3 pumps (inactive) U (NPC or UPC) + 0.5 SSEm 1.1S mb+15. 9.a (NPC or UPC), 0.5 SSE Not applicable Not applicable Not applicable E EPC m 1.1S mb+15. 9.a EPC F NPC + SEE + DSL m 1.2S mb+15. 9.b NPC + SEE + DSL Class 2 & 3 pumps (inactive) U (NPC or UPC) + 0.5 SSEm 1.1S mb+15. 10.a (NPC or UPC), 0.5 SSE m = 1.1S Code case 1636, NC/ND3423 Reflects industry

position E EPC m 1.1S mb+15. 11/10.a EPC,0.5 SSE + transient (a) (c) [see note (b)]

F NPC + SSE + DSL m 1.1S mb+15. 10.a NPC + SSE + DSL m = 1.2S Class 2 & 3 valves U (NPC or UPC) + 0.5 SSE1.1 Pr 11.a (NPC or UPC), 0.5 SSE m = 1.1S Code case1636, Equally conservative (inactive) E EPC 1.1 Pr 11.a EPC,0.5 SSE + transient (c)NC/ND3621 F NPC + SSE + DSL 1.2 Pr 11.b NPC + SSE + DSL m = 2.0S [see note (b)]

Class 2 & 3 valves U (NPC or UPC) + 0.5 SSE1.0 Pr 12.a (NPC or UPC), 0.5 SSE m = 1.1S Code case1636, Equally conservative (active) E EPC 1.0 Pr 11/ 12.a EPC,0.5 SEE + transient (a)NC/ND3621 (e)

F NPC + SSE + DSL 1.0 Pr 12.a NPC + SSE + DSL m = 1.2S (c)[see note (b)]

C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007 1.8-43 COMPARISON WITH REGULATORY GUIDE 1.48 (Continued)

NOTES Numerical indicators (e.g., 1/) in the regulatory guide portion of the table correspond to the footnotes of Regulatory Guide 1.48. Alphabetical indicators in CGS portion of table (or comparative column) correspond to the following:

aIn addition to compliance with the design limits specified, assurance of operability under all design loading combinations shal l be in accordance with Section 3.9.3.2. b Referenced paragraphs of code currently in course of preparation.

cThe design limit for local membrane stress intensity or primary membrane plus primary bending stress intensity is 150% of that allowed for general membrane (except as limited to 2.4S for inactive components under faulted condition). See Section 3.9.5.2. d Not used.

eInactive limits may be used since operability will be demonstrated in accordance with Section 3.9.3.2. fWhen selecting plant events for evaluation, the choice of the events to be included in each plant condition is selected based o n the probability of occurrence of the particular load combination. The combination of loads are those identified in Table 3.9-2.

LEGEND: UPC = upset plant conditions NPC = normal plant conditions EPC = emergency plant conditions DSL = dynamic system loading SSE = safe shutdown earthquake C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-07-044 1.8-44 DELETED Contents of Regulatory Guide 1.49, Revision 1, December 1973

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-45 Regulatory Guide 1.50, Revision 0, May 1973 Control of Preheat Temperature for Welding of Low-Alloy Steel Regulatory Guide Intent

This guide delineates preheat temperature control requirements and welding procedure qualifications supplementing those in ASME Sections III and IX.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment:

The use of low-alloy steel is restricted to the reactor pressure vessel. Other ferritic components in the RCPB are fabricated from carbon steel materials.

Preheat temperatures employed for welding of low-alloy steel meet or exceed the requirements of ASME Section III. Components were either held for an extended time at preheat temperature to ensure removal of hydrogen, or preheat was maintained until postweld heat treatment. The minimum preheat and maximum interpass temperature were specified and monitored.

All welds were nondestructively examined by radiographic methods. In addition, a supplemental ultrasonic examination was performed.

By meeting and/or exceeding the recommendation of the ASME Code, the intent of the regulatory guide is satisfied even though the design was significantly developed prior to issuance of the specific guide wording.

Specific Evaluation Reference

See Section

5.2.3. Similar

Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-46 Regulatory Guide 1.53, Revision 0, June 1973 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems Regulatory Guide Intent

Regulatory Guide 1.53 requires that protection systems meet the requirements of Section 4.2 of IEEE 279-1971, which is also required by ANSI-N 42.7-1972 in that any single failure within the protection systems shall not prevent proper protective action at

the system level when required. This guide provides guidance on an acceptable method of complying with this requirement.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in full compliance with this regulatory guide.

General Compliance or Alternate Approach Assessment:

Compliance is achieved by specifying, designing, and constructing the engineered safeguards systems to meet the single failure criterion, Section 4.2 of IEEE 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, and IEEE 379-1972, IEEE Trial-Use Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection Systems.

This regulatory guide applies to the following NSSS supplied protection systems:

reactor protection system (RPS), ECCS, and PCRVICS.

The reactor protection system has separate and redundant instrument channels, logic, and actuation circuits to ensure that the single failure criterion is met. The PCRVICS is

similarly designed.

The ECCS is divided into the ADS, HPCS, LPCS and RHR (LPCI) which meets the

single failure criterion on a network basis.

Specific Evaluation Reference

See Sections 7.2.2.2 and 7.3.2.1.2. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-47 Regulatory Guide 1.54, Revision 0, June 1973 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear

Power Plants.

Regulatory Guide Intent

This guide describes an acceptable method of complying with QA requirements for

protective coatings.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulations and the NRC regulatory guide or

NRC-approved alternate position.

General Compliance or Alternate Approach Assessment:

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation Reference

Information was provided at the PSAR stage.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-48 Regulatory Guide 1.56, Revision 0, June 1973 Maintenance of Water Purity in Boiling Water Reactors Regulatory Guide Intent

This guide describes an acceptable method of implementing GDC 13, 14, 15, and 31 with regard to minimizing the probability of corrosion-induced failure of the RCPB in BWRs by maintaining acceptable purity levels in the reactor coolant and acceptable instrumentation to determine the condition of the reactor coolant.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

Materials in the primary system are primarily Type 304 stainless steel and Zircaloy cladding. The reactor water chemistry limits have been established to provide an environment favorable to these materials. Design and Licensee Controlled Specifications (LCS) limits are placed on conductivity and chloride concentrations.

Operationally, the conductivity is limited because it can be continuously and reliably measured and gives an indication of abnormal conditions and the presence of unusual materials in the coolant. Chloride limits are specified to prevent stress corrosion

cracking of stainless steel.

The water quality requirements are further supported by GE topical report

NEDO-10899.

Specific Evaluation Reference

See Section

5.2.3. Similar

Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-49 Regulatory Guide 1.58, Revision 0, August 1973 Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel

Regulatory Guide Intent

This guide describes an acceptable method of complying with the NRCs regulations on

qualification of nuclear power plant inspection, examination and testing personnel.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulations and the NRC regulatory guide or

NRC-approved alternate position.

General Compliance or Alternate Approach Assessment:

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation Reference

Information was provided at the PSAR stage. Compliance is discussed in the OQAPD.

Similar Application Reference

Similar application has not been used in other plants.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-50 Regulatory Guide 1.60, Revision 1, December 1973 Design Response Spectra for Seismic Design of Nuclear Power Plants.

Regulatory Guide Intents

This guide delineates procedures for defining response spectra for designing Seismic Category I structures, systems, and components.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

The input loadings for the seismic analysis of the CGS plant structures were given in terms of response spectra based on data available on earthquake acceleration time

history records which was accepted industry practice at the time of the CGS design.

This method was acceptable to the NRC prior to the issuance of this regulatory guide

because no other guidance was available.

Specific Evaluation Reference

See Section 3.7.1.1. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-51 Regulatory Guide 1.61, Revision 0, October 1973 Damping Values for Seismic Design of Nuclear Power Plants Regulatory Guide Intent

This guide delineates damping values that should be applied to modal dynamic analysis

of Seismic Category I elements.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

The damping values used in the seismic analysis conform to the data available on this at

the time the analysis was performed which was the practice accepted by industry and

the NRC at the time of the CGS design.

The values used in Table 3.7-1 are less than those given by the regulatory guide. The calculated responses are therefore conservative.

Specific Evaluation Reference

See Section 3.7.1.3. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-52 Regulatory Guide 1.62, Revision 0, October 1973 Manual Initiation of Protective Actions.

Regulatory Guide Intent

Regulatory Guide 1.62 requires that manual initiation of each protective action at the system level be provided, that such initiation accomplishes all actions performed by

automatic initiation, and that protective action at the system level go to completion once manually initiated. In addition, manual initiation should be by switches readily accessible in the control room, and a minimum of equipment should be used in common with automatically initiated protective action.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility

is in full compliance with this regulatory guide.

General Compliance or Alternate Approach Assessment:

Means are provided for manual initiation of primary containment and reactor vessel isolation control system (NSSS only), ECCS, and reactor protection system scram at the system level through the use of armed push buttons, as described below:

Action Initiated Number ofSwitches Primary containment and reactor

vessel isolation (NSSS Only) Four, two in Division 1 and two in

Division 2 ADS Four, two in Division 1 and two in Division 2 HPCS One switch in Division 3 RHR (loop A)/LPCS One switch in Division 1 RHR (loop B)/RHR (loop C)

One switch in Division 2 Reactor protection system (SCRAM) Four, two in Division 1 and two in

Division 2

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-53 Operation of these switches accomplishes the initiation of all actions performed by the automatic initiation circuitry.

The amount of equipment common to both manual and automatic initiation of the above

function is kept to a minimum through implementation of manual activation as close as possible to the final devices actuators (relays, scram contractor) of the protection system. No failure in the manual, automatic or common portions of the protection system will prevent initiation of a given function by manual or automatic means.

Manual initiation of any of the above functions, once initiated, goes to completion as required by IEEE 279-1971, Section 4.16.

Specific Evaluation Reference

See Sections 7.2.2.3 and 7.3.2.1.3. Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-54 Regulatory Guide 1.64, Revision 2, June 1976 Quality Assurance Requirements for the Design of Nuclear Power Plants

Regulatory Guide Intent

This guide describes an acceptable method of complying with the NRCs QA

requirements for the design of the nuclear power plants.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulations and the NRC regulatory guide or

NRC-approved alternate position.

General Compliance or Alternate Approach Assessment:

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation Reference

Information was provided at the PSAR stage. Compliance is discussed in the OQAPD.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-55 Regulatory Guide 1.65, Revision 0, October 1973 Materials and Inspection for Reactor Vessel Closure Studs.

Regulatory Guide Intent

Regulatory Guide 1.65 defines acceptable materials and testing procedures with regard

to reactor vessel closure stud bolting for light-water-cooled reactors.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

The reactor pressure vessel closure studs are SA-540 Grade B23 or 24 (AISI4340) and have a maximum ultimate tensile strength of 170 ksi. Additionally, specified bolting

material must have Charpy V notch impact properties of 45 ft-lb minimum with 25 mils lateral expansion. Nondestructive examination before and after threading is specified to be in accordance with subarticle NB-2580 ASME Section III, which complies with

regulatory position C.2. Subsequent to fabrication, the studs are manganese phosphate coated and are lubricated with a graphite/alcohol or a nickel powder base lubricant.

In relationship to regulatory position C.2.b, the bolting materials were ultrasonically examined after final heat treatment and prior to threading, as specified. The specified

requirement for examination according to ASME Section II Recommended Practice SA-388 was complied with. The specific procedures approved for use in practice are

judged to ensure comparable material quality and, moreover, are considered adequate on the basis of compliance with the applicable requirements of ASME Section III paragraph NB-2585.

Additionally, straight beam examination was performed on 100% of cylindrical

surfaces, and from both ends of each stud using a 3/4 maximum diameter transducer.

In addition to the code required notch, the reference standard for the radial scan contained a 0.5-in. diameter flat bottom hole with a depth of 10% of the thickness, and the end scan standard contained a 0.25-in. diameter flat bottom hole 0.5-in. deep.

Also, angle beam examination was performed on the outer cylindrical surface of nuts C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-021 1.8-56 and washers per ASME SA-388 in both an axial and circumferential direction. Any indication greater than the indication from the applicable calibration feature is unacceptable. A distance-amplitude correction curve per NB-2585 is used for the longitudinal wave examination. Surface examinations were performed on the studs and nuts after final heat treatment and threading, as specified in the Regulatory Guide, in

accordance with NB-2583 of ASME Code Section III, 1971 Edition through

November 1971 Addenda.

In relationship to regulatory position C.3, GE practice allows exposure of stud bolting surfaces to high purity fill water; nuts and washers are stored dry during refueling.

In relationship to regulatory position C.4, ASME Section XI and Appendix VIII guidance is followed, as implemented in the In-Service Inspection Program, to monitor the structural integrity of the stud bolting and ability to perform their function.

Specific Evaluation

Reference:

See Section 5.3.1.7. Similar Application

Reference:

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-57 Regulatory Guide 1.66, Revision 0, October 1973 Nondestructive Examination of Tubular Products.

Regulatory Guide Intent

This guide describes a method of implementing requirements acceptable to NRC regarding nondestructive examination requirements of tubular products used in the

RCPB. Applicable Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

Wrought tubular products were supplied in accordance with applicable ASTM/ASME material specifications. These specifications require a hydrostatic test on each length of tubing. Additionally, the specification for the tubular product used for CRD housings specified ultrasonic examination to paragraph NB-2550 of ASME Code Section III.

These RCPB components met the requirements of ASME Codes existing at time of placement of order which predated Regulatory Guide 1.66. At the time of the placement of the orders, 10 CFR 50, Appendix B requirements and ASME code requirements assured adequate control of quality for the products.

This regulatory guide was withdrawn on September 28, 1977, by the NRC because the additional requirements imposed by the guide were satisfied by the ASME Code

Section III.

Specific Evaluation Reference

See Sections 4.5.2.3 and 5.2.3.3. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-58 Regulatory Guide 1.67, Revision 0, October 1973 Installation of Overpressure Protection Devices

Regulatory Guide Intent:

This regulatory guide describes a method acceptable to the NRC staff for implementing GDC 1 with regard to the design of piping for safety valve and relief valve stations which have open discharge systems with limited discharge pipes and which have inlet piping that neither contains a water seal nor is subject to slug flow of water on discharge of the valves.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified RHR shutdown suction line thermal relief piping is located between the containment isolation valves. However, the intent of the regulatory guide does not

apply due to the very short duration and small discharge of the thermal relief function.

General Compliance or Alternate Approach Assessment

This regulatory guide is not considered to be applicable to this piping due to the small size and very short operation time of the valve (0.75 in. x 1 in.). The only purpose of the valve is to relieve the excess pressure caused by the difference of thermal expansion

between the pipe and the water contained between the containment isolation valves.

Specific Evaluation Reference

See Section 3.9.3.1.14. Similar Application Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-59 Regulatory Guide 1.68, Revision 0, November 1973 Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors

Regulatory Guide Intent

Regulatory Guide 1.68 describes the requirements for the initial startup test programs.

This regulatory guide is applicable to such activities as precritical tests and low-power tests. Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the

alternate approach cited.

General Compliance or Alternate Approach Assessment

The following discussion describes the alternate acceptable approaches for specific

conformance to this regulatory guide.

The format of the CGS test procedures is different from that of the guide, but since the content specifies the required elements, the procedures are in compliance.

The reference sections refer to those of the regulatory guide. Those sections not listed

are in compliance.

Section C.2.b: Operational limitations for the protection of public health and safety are included in the Technical Specifications for the plant. The General Electric startup instructions contain notes of caution which supplement the Technical Specifications.

The Technical Specifications should be the instrument for describing operational (including testing) limitations. Therefore, the identification of safety precautions in test procedures should be limited to those items which, if not observed, could lead to reduction of system safety performance below expected levels and not the minor

procedural and test details which would not cause such a reduction.

Section C.2.c

The generic simulation test appearing in Chapter 14 should appear by reference in preoperational and initial startup test programs where onsite full

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-60 simulation tests are not possible. The guide wording would change to ... less than full simulation should be provided or referenced for test where full...

Appendix A, Section C.2.h: The comparison of critical control rod pattern with predicted patterns (Appendix A, Section C.2.d) provides required knowledge of effective overall rod worth. Individual control rod calibrations cannot be performed in a meaningful manner in a large multirodded BWR. Therefore, this part of the guide is

not applicable to BWRs.

Appendix A, Section C.2.i: The functional requirement of the reactor head cooling system design is required at operating pressures less than or equal to 135 psig.

Therefore, for this paragraph to be applicable (135 psig) should be part of last

sentence.

Appendix A, Section D.2.a

The high-pressure coolant injection (HPCI) has been replaced by an HPCS system. Due to the configuration of the sprays directly on the core, this system cannot be operated at power. The HPCS injection/core spray is demonstrated during the preoperational test program.

Appendix A, Section D.2.b: Friction tests are performed on four drives at rates pressure.

Appendix A, Section D.2.f: It is necessary to make more than two calibrations and, therefore, it is not appropriate to limit the test to 50% and 100% power levels.

Appendix A, Section D.2.g: At least six chemical analyses of fluid system are necessary; therefore, the limitations of 25%, 50%, 75%, and 100% are not

appropriate.

Appendix A, Section D.2.1: Since this plant design does not include an emergency condenser, this section is not appropriate.

Appendix A, Section D.2.n: Control rod calibration in a large multirodded BWR has not been found to provide meaningful data. Any safety-related problems associated with control rods would be discovered during safety related testing, and therefore, this section is not appropriate.

Appendix A, Section D.2.p: Since the main steam valve function tests are conducted at a minimum of six power and flow conditions, the limitations of 25%, 50%, and 75%

are not appropriate.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-61 Appendix A, Section D.2.s and t: Turbine trip and generator trip have essentially the same effect on the reactor and safety related system actuation. Sections D.2.s and D.2.t should be combined into one test.

Appendix A, Section D.2.y: Minimum critical heat flux ratio (MCHFR) is an obsolete limit that has been replaced with minimum critical power ratio (MCPR). Core

performance evaluation tests must be performed at every test condition.

Appendix A, Section D.2.aa: Comparison tests are made throughout the test program, and therefore, limitations of 25%, 50% and 100% are not appropriate.

Appendix C, Section B.2.d: Functionally testing the associated control rod immediately following installation of each fuel cell is not appropriate. Functional testing of all control rods after fuel loading and prior to startup to critical procedures is applicable.

Appendix A, Section A.5.a: The demonstration of water injection for a LOCA is an ECCS test. Therefore, demonstration of water injection for a loss-of-coolant accident is not within the scope of the reactor coolant makeup system test.

Appendix A, Section C.2.c: The calibration of intermediate range monitor with power is not meaningful due to local control rod effects.

Appendix A, Section D.2.w

Feedwater pump trip should be performed to check recirculation pump runback.

Appendix C, Section B.1.b

Poison curtains are not applicable since they are not used in this plant.

Appendix C, Section B.2.a

Poison curtains are not applicable.

Appendix C, Section B.3.c: The insertion of locked control rods is excluded in any withdrawal sequence.

Appendix D, Section D.2.0: The rod pattern exchange is not a part of the Startup Power Ascension Program since it does not involve the approach of any safety margin or operating limit. The rod pattern exchange procedure at power is part of the Nuclear Performance Evaluation Procedure and will be performed during the fuel cycle as necessary. The simultaneous trip of both recirculation pumps is not performed at 100%

of rated power. The analysis of this event (see Section 15.3.1) indicates there is no decrease in the MCPR and therefore, it does not involve the approach of any safety

margin or operating limit.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-62 Specific Evaluation Reference

See Section 14.2. Similar Application Reference

Similar application was used for Brunswick 1 and Browns Ferry 3.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-029 1.8-63 Regulatory Guide 1.70, Revision 2, September 1975

Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants

Regulatory Guide Intent:

This guide describes the minimum acceptable requirements for format and content of

Safety Analysis Reports.

Application Assessment:

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in full compliance with this regulatory guide or through the incorporation of the NRC approved alternate approach cited.

General Compliance or Alternate Approach Assessment:

The NSSS scope of supply inputs include all the appropriate scope responsibilities and information required in Regulatory Guide 1.70, Revision 2, in both format and content, except as described below. Appendix A of NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR II) (most recent approved revision referenced in the COLR), provides a road map for incorporating nuclear fuel design and analysis characteristics described in GESTAR II into the FSAR. GESTAR II is consistent with Regulatory Guide 1.70, Revision 3.

Specific Evaluation

Reference:

For Regulatory Guide 1.70, Revision 2, see NSSS scope of supply portions of this FSAR. For Regulatory Guide 1.70, Revision 3, see Sections 4.1 , 4.2 , 4.3 and 4.4. Similar Application

Reference:

Similar application was used for Grand Gulf 1 and 2 and Susquehanna 1 and 2.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-64 Regulatory Guide 1.71, Revision 0, December 1973 Welder Qualification for Areas of Limited Accessibility

Regulatory Guide Intent

Regulatory Guide 1.71 requires that weld fabrication and repair for wrought low-alloy

and high-alloy steels or other materials such as static and centrifugal castings and

bimetallic joints should comply with fabrication requirements of Section III and Section IX of the ASME B&PV Code. It also requires additional performance qualifications for welding in areas of limited access.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

All ASME Section III welds were fabricated in accordance with the requirements of

Section III and IX of the ASME B&PV Code. There are few restrictive welds involved in the fabrication of BWR components. Welder qualification for welds with the most restrictive access was accomplished by mock-up welding. Mock-ups were examined

with radiography or sectioning.

All reactor pressure boundary welding was performed in accordance with ASME

Section IX. Reactor internal component welding was performed in accordance with

ASME Section IX or appropriate AWS requirements.

Specific Evaluation Reference

See Section

5.2.3. Similar

Application Reference

Similar application was used for Zimmer and LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-65 Regulatory Guide 1.73, Revision 0, January 1974 Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants.

Regulatory Guide Intent

Regulatory Guide 1.73 endorses the requirements of IEEE 382-1972, Trial-Use Guide for Type Test of Class 1 Electric Valve Operators for Nuclear Power Generating Station. Regulatory position stipulations are also included.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility

is in full compliance with this regulatory guide.

General Compliance or Alternate Approach Assessment

This regulatory guide is applicable to the recirculation system gate valve and the HPCS

injection valve motor operators.

These valve operators have been tested in accordance with the test sequence outlined in Section 4.5.2 of the IEEE 382-1972. The qualifying tests have been made under environmental conditions (temperature, pressure, humidity, radiation) that are at least as severe as those that the valve operator will be exposed to during and following a

DBA (LOCA).

Specific Evaluation Reference

See Section 3.11. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-66 Regulatory Guide 1.74, Revision 0, February 1974 Quality Assurance Terms and Definitions

Regulatory Guide Intent

This guide identifies quality assurance terms and acceptable definitions.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulations and the NRC regulatory guide or

NRC-approved alternate position.

General Compliance or Alternate Approach Assessment

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation Reference

Information was provided at the PSAR stage. Compliance is discussed in the OQAPD.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-67 Regulatory Guide 1.75, Revision 0, February 1974 Physical Independence of Electrical Systems

Regulatory Guide Intent

This guide presents a detailed method of ensuring physical independence of electric systems, including requirements of preparation, identification, and isolation.

Application Assessment

Assessed capability in design

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment

When evaluating the applicability of Regulatory Guide 1.75 and its attendant IEEE Standard (IEEE-384-1971), consideration should be given to the fact that design was

significantly developed prior to their issuance.

The following is a point-by-point definition of the implementation of IEEE-384 as modified by Regulatory Guide 1.75 for the CGS plant. The numbers and titles in the

following see those of IEEE-384.

1. Scope Compliance with scope.

2. Purpose

Compliance with purpose.

3. Definitions

All definitions apply including Regulatory Guide 1.75 except for small nomenclature aspects in C.1 and C.2 associated within floor sections.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-68 4. General Separation Criteria

4.1 Required

Separation

4.2 Equipment

and Circuits Requiring Separation The equipment and circuits requiring separation are determined and delineated early in the plant design. Distinctive identification of those

equipment and circuits were not provided on specifically noted documents and drawings but the documents and drawings are identified as applying to the protection systems.

4.3 Methods

of Separation Barriers are used to separate divisional devices and wiring. Safety system logic is implemented with relay coil to relay contact separation of multidivisional and nondivisional signals. Distance separation was provided to the extent feasible at manufacturing time. These served the

purpose or intent of requirements at that time.

4.4 Compatibility

with Mechanical Systems The Class 1E equipment and circuits are specified to be located so that a failure in the mechanical systems served by the Class 1E systems does not disable redundant portions of the Class 1E systems.

• 4.5 Associated Circuits Associated circuits are treated as non-Class 1E circuits and are separated

to the extent that good electrical isolation is assured. This assurance was provided without Class 1E isolators. Some physical separation is provided.

4.6 Non-Class 1E Circuits

4.6.1 Separation

from Class 1E Circuits Same as 4.5 response above.

• Information on compliance of actual installation is provided in Section 1.8.3.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-69 4.6.2 Separation from Associated Circuits Same as 4.5 response above.

5. Specific Separation Criteria

5.1 Cables

and Raceways To the extent that the 5.1 series of subparagraphs might be used to critique the power generation control complex (PGCC) equipment, the physical reality of the floor sections is obviously

not recognized in the IEEE-384 test. However, the floor sections are inherently in accordance with the design concepts stated in

these subparagraphs and therefore comply on that basis.

5.2 Standby

Power Supply Comply as applied to the Division 3 HPCS Diesel Generator.

• 5.3 DC System Comply as applied to the Division 3 HPCS Diesel Generator.

• 5.4 Distribution System Comply as applied to the Division 3 HPCS Diesel Generator.

• 5.5 Containment Electrical Penetrations Not in NSSS scope of supply.

5.6 Control

Switch Boards

5.6.1 Location

and Arrangement Class 1E equipment and circuits are located on separate control switchboards or where operationally necessary on

a single control switchboard.

• Division 1 and 2 power compliance is provided in Section 1.8.3. The control room structure and location as well as local control switchboard location is discussed in Section 1.8.3.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-70 5.6.2 Internal Separation Most of the devices requiring separation are separated by barriers. With several divisions in one panel, and for relays which must accept multidivisional signals, 6-inch separation is impossible. Therefore, separation is done on a best effort approach. Design has used the relay coil to

relay contact separation to comply with the regulatory

guide. 5.6.3 Internal Wiring Identification Panel internals wiring is not color-coded, but wires are marked with their respective Connection Diagram identify

at each point of termination.

5.6.4 Common

Terminations Relay coil to relay contact separation has been used.

5.6.5 Non-Class 1E Wiring Electrical isolation is provided, though not necessarily with Class 1E isolators. Some physical separation is provided.

5.6.6 Cable

Entrance Not in NSSS scope of supply.

5.7 Instrumentation

Cabinets Compliance

5.8 Sensors

and Sensor to Process Connections Compliance

5.9 Actuated

Equipment Not in NSSS scope of supply.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 1.8-71 Specific Evaluation Reference

See Section 8.3.1.4.2.7 Similar Application Reference

Application of this regulatory guide is plant unique due to NRC agreements during the

various stages of licensing and scope of responsibility of design and engineering necessary to comply with the NRC interpretation. Therefore reference plants cannot be cited.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-004 1.8-72 Regulatory Guide 1.84

Design, Fabrication, and Materials Code Case Acceptability, ASME Section III

Regulatory Guide Intent

This guide lists all Section III Code Cases that the NRC has approved for use. It is updated on a regular basis to reflect the changes to the ASME Code Cases and the current position of the NRC on acceptability for use. The guide contains tables that detail the NRC acceptance requirements for current, annulled, and superseded Code Cases. Code Cases that the NRC determined to be unacceptable are listed in Regulatory Guide 1.193, ASME Code Cases Not Approved for Use.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

The current version of the Regulatory Guide is utilized to determine acceptable Code Cases for all new and existing plant applications. The FSAR does not track individual Code Cases and revision numbers. Not all acceptable Code Cases listed in the regulatory guide are used. The Code Cases that are utilized for Columbia are referred to in the plant design/installation documentation.

General Compliance or Alternate Approach Assessment:

Code Cases are utilized in accordance with the requirements of the regulatory guide provisions for acceptance.Section III Code Cases that are not yet endorsed may be utilized via submittal to the NRC for approval in accordance with the regulatory guide. The plant scope of supply is in full compliance with this regulatory guide.

Specific Evaluation Reference

See Section

3.2. Similar

Application Reference

None.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-004 1.8-73 Regulatory Guide 1.85, Revision 31, 1998

• Code Case Acceptability ASME Section III Materials

Regulatory Guide Intent:

This guide provides a list of ASME materials code cases that have been generically

approved by the NRC.

Code cases on this list may be used until annulled. Annulled cases are considered active for equipment that has been contractually committed to fabrication prior to the

annulment.

This guide and later revisions require NRC approval of code cases for Class 1, 2, and 3

components.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment:

The GE procedure is to obtain NRC approval of code cases on Class 1 components only. NRC approval of Class 2 and 3 code cases was not required by 10 CFR 50.55(a).

All Class 2 and 3 equipment has been designed to ASME Code or ASME approved Code Cases. This provision together with quality control requirements provide adequate safety equipment functional assurances.

Specific Evaluation Reference

See Section 5.2.1.

• Regulatory Guide 1.85 was withdrawn in 2004. See Regulatory Guide 1.84 for NRC acceptance of current Materials Code Cases.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-004 1.8-74 Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-75 Regulatory Guide 1.88, Revision 2, October 1976 Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records.

Regulatory Guide Intent

This guide describes an acceptable method of complying with the NRCs regulations for

collection, storage, and maintenance of quality assurance records.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

The identified BWR Quality Assurance Program used in this facility reflects compliance with the provisions of NRC regulations and the regulatory guide or NRC-approved

alternate position.

General Compliance or Alternate Approach Assessment:

Reference compliance assessment for Regulatory Guide 1.28.

Specific Evaluation

Reference:

Information was provided at the PSAR stage. Compliance is discussed in the OQAPD.

Similar Application Reference

Similar application has not been used for other projects.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-020 1.8-76 Regulatory Guide 1.89, Revision 1, June 1984

Qualification of Class 1E Equipment for Nuclear Power Plants

Regulatory Guide Intent:

Regulatory Guide 1.89 Rev. 1 endorses both the requirements and recommendations of

IEEE 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear

Power Generating Stations. Additional regulatory position stipulations are also included.

Compliance or Alternate Approach Statement:

CGS complies with this regulatory guide for equipment requiring environmental qualification procured after February 22, 1983.

General Compliance or Alternate Approach Assessment:

For equipment requiring environmental qualification installed prior to February 22, 1983, CGS follows the guidance in NUREG-0588 Cat II.

In view of the NRC Memorandum and Order (CLI-80-21), dated May 23, 1980, all environmental qualifications of Class 1E equipment within the NSSS scope of supply was reevaluated for compliance with NUREG-0588, Category II. Where significant deviation from those guidelines was found in specific equipment qualifications, additional testing and/or analysis was performed to demonstrate the adequacy of the

equipment to perform its safety-related function.

Specific Evaluation

Reference:

Delineation of the degree of compliance is contained in Section 3.11.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-77 Regulatory Guide 1.92, Revision 1, February 1976 Combination of Modes and Spatial Components in Seismic Response Analysis.

Regulatory Guide Intent

This guide describes methods acceptable to the NRC for combining the values of the response spectrum nodal dynamic analysis and in combining maximum values (in case of time history dynamic analysis) or the representative maximum values (in case of spectrum dynamic analysis).

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement:

Identified NSSS scope of supply analysis, design, and/or equipment used in this facility is in compliance with the intent of this regulatory guide through the incorporation of the alternate approach cited.

General Compliance or Alternate Approach Assessment:

Three Components of Earthquake Motion

Response Spectrum Method

The use of three components of earthquake motion was not a design basis requirement

of the construction permit for this plant. The total seismic response is predicted by combining the response calculated from analyses due to one horizontal and one vertical seismic input. For this case, where the response spectrum method of seismic analysis is used, the basis for combining the loads from the two analyses is given as follows:

a. The peak of the different modes for the same earthquake excitations do not occur at the same time,

b. The peak responses of a particular mode due to earthquake excitations from different directions do not occur at the same time, and

c. The peak stresses due to different modes and due to different excitations may not occur at the same location nor in the same direction.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-78 To implement the above, the two translation components of earthquake excitations are combined by summing the absolute sum of all responses of interest (e.g., strain, displacement stress, moment, shear, etc.) from seismic motion, the one horizontal (x or z) and one vertical direction (y), i.e., x+y or y+z . The design is made for the larger of the two sums x+y or y+z. Time History Method

The algebraic sum of contributions (to displacements, loads, stresses, etc.) due to the two earthquake components are calculated for each natural mode for each time interval of analysis. The time interval should be less than or equal to 0.2 of the smallest period of interest. The maximum values of all time intervals are the design displacements, accelerations, loads, or stresses.

It is concluded that the above method adequately demonstrates the integrity of the Seismic Category I subsystems and was found acceptable as a basis of current operating

BWR plants.

Combination of Modal Responses

When the response spectra method of modal analysis is used, all modes are combined by the square root of the sum of the squares (SRSS) described as follows:

The SRSS combination of modal responses is defined mathematically as

RR i i n2 1 where R = Combined response

R i = Response in the i th mode n = Number of modes considered in the analysis

Closely spaced modes are not accounted for as required by the guide because the design was significantly developed prior to issuance of the guide.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-79 Specific Evaluation Reference

See Sections 3.7.3.6 and 3.7.3.7. Similar Application Reference

Similar application was used for LaSalle.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-80 Regulatory Guide 1.99, Revision 2, May 1988 Radiation Embrittlement of Reactor Vessel Materials

Regulatory Guide Intent:

This regulatory guide provides guidance for the prediction of irradiation damage of the reactor vessel belt line materials for the life of the vessel. This information is used to develop the pressure/temperature limit curves for the reactor pressure vessel based on material chemistry and end-of-life neutron exposure.

Application Assessment:

Assessed capability in design.

Compliance or Alternate Approach Statement:

The reactor pressure vessel pressure/temperature limit curves are in full compliance

with the identified requirements in the regulatory guide.

General Compliance or Alternate Assessment:

Compliance is achieved by using a calculated end-of-life fluence for the CGS reactor vessel to evaluate the material damage due to this fluence. This information is used to predict the end-of-life NDT temperature for the limiting belt line material for the vessel. Using linear elastic fracture mechanics, the requirements of Welding Research Council Bulletin 175, the Standard Review Plan, and the requirements of Regulatory Guide 1.99, Revision 2, the pressure/temperature limit curves were developed for CGS. These curves will be used to evaluate the predictions determined by the regulatory guide until the submittal of new curves that incorporate the results of the

surveillance capsule test data.

Specific Evaluation

Reference:

See Sections 5.3.1.5.2.1 through 5.3.1.5.2.6 and the Technical Specifications.

Similar Application

Reference:

Similar application is used on all reactor vessels.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-011 1.8-81 Regulatory Guide 1.100, Revision 1, August 1977

Seismic Qualification of Electric Equipment for Nuclear Power Plants

Regulatory Guide Intent

Regulatory Guide 1.100 endorses both the requirements and recommendations of IEEE 344-1975, IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, when such qualification is performed in conjunction with Regulatory Guide 1.89, and subject to the regulatory position stipulations.

Compliance or Alternate Approach Statement

General Compliance or Alternate Approach Assessment

All Class 1E equipment seismic qualifications are evaluated against the requirements set forth within IEEE 344-1975 as clarified in Section 3.10.1.2. The evaluations are documented and demonstrated adequacy of the methods and results of the qualifications as equal or conservative to the requirements of IEEE 344-1975. This qualification documentation includes evaluation of seismic and hydrodynamic load combinations.

Specific Evaluation Reference

See Section 3.10 and WNP-2 Dynamic Qualification Report for Safety-Related Equipment, dated September 1982.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-82 Regulatory Guide 1.145, Revision 1, November 1982/February 1983 Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear

Power Plants

Regulatory Guide Intent This guide provides acceptable methodology to determining site-specific off-site air dispersion factors (/Q) for assessing the potential offsite radiological consequences of postulated accidental releases of radioactive material to the atmosphere.

Application Assessment Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and equipment used in this facility is

in full compliance with the regulatory guide.

General Compliance or Alternate Approach Assessment Two of the procedures contained in the PAVAN code were implemented. The procedures were run with the desert sigma and with the Pasquill-Gifford sigma enabled. The most conservative /Q values were used in the accident analysis.

Specific Evaluation Reference

See Section 2.3 and Chapter 15.0.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-83 Regulatory Guide 1.183, Revision 0, July 2000 Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear

Power Reactors

Regulatory Guide Intent

This guide provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. This guide establishes an acceptable alternative source term (AST) and identifies the significant attributes of other ASTs that may be found

acceptable by the NRC staff. This guide also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and equipment used in this facility is

in compliance with this regulatory guide or through the incorporation of the NRC

approved alternate approach cited.

General Compliance or Alternate Approach Assessment

This regulatory guide is applicable to the analyses for the FSAR. The Columbia analysis methods and assumptions (see Energy Northwest, Columbia Generating Station Alternative Source Term, CGS-FTS-0168, Revision 0, August 2007) conform to position of this Regulatory Guide with the following specific considerations.

[Guide Section 3.4] Table 5 of the regulatory guide lists the elements in each radionuclide group that should be considered in design basis analyses. The intent of the guidance is met by an alternate approach. The Columbia analyses consider 66 nuclides consisting of 60 identified as being potentially important contributors to TEDE in NUREG/CR-4691 plus seven additional noble gas isotopes and Ba-137m.

[Guide Section 4.3] Columbia conforms with guide section 4.3 with the exception that the TID-14844 source term continues to be used as the radiation dose basis for

equipment qualification.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-84 [Guide Section 3.3 of Appendix A] The intent of the guidance is met by the conservative approach used in the Columbia analysis. The SRP 6.5.2 model is used.

Elemental iodine is assumed to be removed at the same rate as particulate. The approach of treating elemental iodine as particulate is a conservative representation of

the situation in which some elemental iodine would be removed by diffusion to spray water droplets and some elemental iodine would adsorb onto particulate. A reduction of 10 in iodine removal lambda is taken when 98% of the particulate has been removed.

The method results in a conservative dose.

Specific Evaluation Reference

See Chapter 15.4.9 , 15.6.4 , 15.6.5 , 15.7.4. Similar Application Reference

Similar application was used for Grand Gulf and Brunswick.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-85 Regulatory Guide 1.190, Revision 0, March 2001 Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence

Regulatory Guide Intent

This Regulatory Guide has been developed to provide state-of-the-art calculations and measurement procedures that are acceptable to the NRC staff for determining pressure

vessel fluence.

Application Assessment:

Assessed capability in design.

Compliance or Alternate Approach Statement

The methodology for the neutron flux calculation for the CGS reactor vessel conforms to Licensing Topical Report (LTR) NEDC-32983-P-A. In general, the methodology described in the LTR adheres to the guidance in Regulatory Guide 1.190 for neutron

flux evaluation and was approved by the U.S. NRC in the Safety Evaluation Report (SER) for referencing in Licensing submittals.

General Compliance or Alternate Assessment

Reference compliance assessment for Regulatory Guide 1.99.

Specific Evaluation Reference

See Section 4.3.2.8.

Similar Application

Reference:

Similar application is used for Browns Ferry Nuclear Plant, Units 2 and 3, reactor

vessels.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-86 Regulatory Guide 1.194, Revision 0, June 2003 Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments

at Nuclear Power Plants

Regulatory Guide Intent

This guide provides guidance on determining atmospheric relative concentrations (/Q) values in support of design basis control room radiological habitability assessments at nuclear power plants. This guide describes methods acceptable to the NRC staff for determining /Q values that will be used in control room radiological habitability assessments performed in support of applications for licenses and license amendment requests.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

Identified NSSS scope of supply analysis, design, and equipment used in this facility is in compliance with this regulatory guide or through the incorporation of the NRC approved alternate approach cited.

General Compliance or Alternate Approach Assessment

This regulatory guide is applicable to the analyses for the FSAR. The Instantaneous Puff Release alternative method provided by this guide is used to calculate /Q for the Main Steam Line Break accident.

Specific Evaluation Reference

See Section 15.6.4. Similar Application Reference

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-87 1.8.3 BALANCE OF PLANT SCOPE OF SUPPLY EVALUATION

The following evaluations of implementation of regulatory guides are relative to BOP scope of supply. Thus, reference to CGS in the following evaluations is restricted to the BOP scope of

supply portions of CGS. For NSSS scope of supply implementation of regulatory guides, see

Section 1.8.2.

Conformance to the regulatory guides falls under either of the two following categories:

a. Compliance with the guidance set forth in this regulatory guide as described in this FSAR or

b. Compliance with the intent of the guidance set forth in this regulatory guide by an alternate approach.

The second category is based on NRC rules which state:

Regulatory guides are not substitutes for regulations, and compliance with them

is not required. Methods and solutions different from those set out in the guides will be acceptable if they provide a basis for the findings requisite to the assurance or continuance of a permit or license by the NRC.

Regulatory guides and their revisions are addressed in the following.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-88 Regulatory Guide 1.6, Revision 0, March 1971 Independence Between Redundant Standby (Onsite) Power Sources and Between Their

Distribution Systems

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The compliance assessments given below correspond numerically to the Regulatory Positions as enumerated in Section C of Regulatory Guide 1.6, Revision 0.

1. The electrically powered safety loads, both ac and dc, are separated into redundant load groups such that loss of any one group will not prevent the

minimum safety function from being performed.

2. Each ac load group has a connection to the preferred offsite power source and to a standby onsite power source. The standby power sources have no automatic connection to any other redundant load groups.

3. Each dc load group is energized by a battery and battery charger. The battery-charger combination has no automatic connection to any other redundant

dc load group.

4. When operating from the standby sources, redundant load groups and the redundant standby sources are independent of each other.

5. A single generator driven by two prime movers in tandem is the standby power source for the Division 1 and 2 ac load groups. The Division 3 ac load group power is supplied by a single generator driven by a single prime mover.

Specific Evaluation Reference

See Sections 8.1.5.2 , 8.3.1.1.7 , 8.3.1.2.1.3 , 8.3.1.2.1.4 , 8.3.1.3 , 8.3.1.4 , 8.3.2.1.1 , 8.3.2.2.1.2 , 8.3.2.3 , and 8.3.2.4.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-13-061 1.8-89 Regulatory Guide 1.8, Revision 1-R, May 1977

Personnel Selection and Training

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

The Regulatory Position of Regulatory Guide 1.8, Rev. 1-R (May 1977) will be implemented.

Specific Evaluation

Reference:

See Sections 13.1.3 , 13.2.1 , and the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-11-002 1.8-90 Regulatory Guide 1.9, Revision 0, March 1971 Selection of Diesel Generator Set Capacity for Standby Power Supplies.

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

The compliance assessments given below correspond numerically to the regulatory positions as enumerated in Section C of Regulatory Guide 1.9, Revision 0.

1. Both the Division 1 and Division 2 diesel generator sets were selected to have a continuous load rating equal to or greater than the sum of the conservative estimated loads needed to be powered at any one time.

2. The predicted loads on both the Division 1 and the Division 2 diesel generator sets do not exceed the 2000-hr rating of either set, respectively, or 90% of the 30-minute rating of either set, respectively.

3. Predicted loads on Division 1 and Division 2 were verified by tests during preoperational testing.

4. The Division 1 and Division 2 diesel generator sets are capable of starting and accelerating to rated speed, in the required sequence, all the needed engineered safety feature and emergency shutdown loads.

The Division 1 and Division 2 diesel generator sets are within the limits of undervoltage, under-frequency, overspeed and voltage and frequency restoration time limits, set forth in the regulatory guide.

5. The suitability of each diesel generator set of the standby power supply was confirmed by prototype qualification test data and preoperational tests.

The scope of Regulatory Guide 1.9, Revision 0 does not include recommendations for surveillance testing. The surveillance requirements for demonstrating the operability of the diesel generators are consistent with the recommendations of Regulatory Guide 1.9 Revision 3 as described in the Bases for Technical Specification B 3.8.1. Compliance with Regulatory Guide 1.9 Rev. 0, as an acceptable basis for the selection of diesel generator sets of sufficient margin to implement General Design Criterion 17, remains as described herein.

Specific Evaluation

References:

See Sections 8.1.5.2 , 8.3.1.1.7 , and 8.3.1.2.1.3.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-91 Regulatory Guide 1.10, Revision 1, January 1973 Mechanical (Cadweld) Splices in Reinforced Bars of Category I Concrete Structures.

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The requirements of the guide have been included in the appropriate specifications for the project construction. Compliance with the guide is ensured by testing and control procedures and reporting program. The program includes splicing crew qualifications, visual inspection of each splice, tensile testing of splice samples, tensile test frequency

program, and a procedure for evaluating substandard test results. The procedure for

testing and sampling of mechanical splices have been implemented.

Specific Evaluation Reference

See Sections 3.8.3.2 and 3.8.4.2 and Table 3.8-4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-92 Regulatory Guide 1.11, Revision 0, March 1971 Instrument Lines Penetrating Primary Reactor Containment.

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

CGS design includes flow restriction orifices and/or excess flow check valves with

position indication in instrument lines which penetrate primary reactor containment. In the event of an instrument line rupture outside primary containment, the integrity and functional performance of the secondary containment system and its associated filtration systems are maintained.

Specific Evaluation Reference

See Sections 7.1.2.4 and 6.2.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-93 Regulatory Guide 1.12, Revision 1, April 1974 Instrumentation for Earthquakes

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

Triaxial strong-motion accelerographs are installed at appropriate locations to provide data on the seismic input to containment; data on frequency, amplitude, and phase relationship of the seismic response of the containment structure; and data on the seismic input to other Category I structures, systems, and components.

Specific Evaluation Reference

See Section 3.7.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-94 Regulatory Guide 1.13, Revision 1, December 1975 Spent Fuel Storage Facility Design Basis

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

A controlled leakage building is provided enclosing the fuel pool. The building is not designed to withstand extremely high winds, but leakage is suitably controlled during refueling operations. The building is equipped with a ventilation and filtration system which is designed to limit the potential consequences of the release of radioactivity specified in Regulatory Guide 1.183 to those requirements set forth in 10 CFR 50.67.

The movement paths of heavy objects such as the reactor pressure vessel head, containment vessel head, and the spent fuel cask are designed not to pass over the spent fuel racks. Furthermore, the reactor building crane and its auxiliary hoist are prevented by means of interlocks from passing over any of the spent fuel pool except the spent fuel cask area. Bypassing of the interlocks is permitted only during fuel handling and storage operations and is administratively controlled.

The fuel pool is designed so that no pipe break will drain water from the fuel pool.

Specific Evaluation Reference

See Section 9.1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-95 Regulatory Guide 1.15, Revision 1, December 1972 Testing of Reinforcing Bars for Category I Concrete Structures

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The requirements of the guide have been included in the appropriate specifications for project construction. Compliance with the guide is assured by the implementation of

qualified testing and control procedures and reporting. Included are qualified control

procedures and reporting for the yield strength and tensile strength tests and

deformation inspections recommended by the guide.

Specific Evaluation Reference

See Sections 3.8.3.2 , 3.8.4.2 , and 3.8.5.2 and Table 3.8-4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-038 1.8-96 Regulatory Guide 1.16, Revision 4, August 1975

Reporting of Operating Information - Appendix A Technical Specifications

Compliance or Alternate Approach Statement:

This regulatory guide was withdrawn in August 2009 and is no longer applicable.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-97 Regulatory Guide 1.17, Revision 1, June 1973 Protection of Nuclear Power Plants Against Industrial Sabotage

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

This information is considered proprietary and is subject to limited distribution. All specifics have been forwarded to the NRC as part of the Energy Northwest proprietary

physical security plan for CGS.

Specific Evaluation Reference

See proprietary physical security plan.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-98 Regulatory Guide 1.18, Revision 1, December 1972.

Structural Acceptance Test for Concrete Primary Reactor Containments

Compliance or Alternate Approach Statement:

This regulatory guide is not applicable since CGS does not have a concrete primary

containment.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-99 Regulatory Guide 1.19, Revision 1, August 1972 Nondestructive Examination of Primary Containment Liner Welds

Compliance or Alternate Approach Statement

This regulatory guide is not applicable since CGS does not have a concrete primary

containment with a steel liner.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-15-062 1.8-100 Regulatory Guide 1.21, Revision 1, June 1974

Measuring, Evaluating, and Reporting of Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear

Power Plants

Compliance or Alternate Approach Statement:

CGS complies with the intent of the guidance established in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment:

The following categories of monitoring systems incorporated into the CGS design fulfill the requirements for monitoring in Regulatory Guide 1.21.

a. Gaseous effluents,

b. Liquid effluents, and

c. Solid Waste.

The above categories of monitoring systems adequately monitor effluent discharge paths

for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

Columbia Generating Station complies with Section C.11.b (Quality Controls) requirements for blind duplicate analysis by an alternate approach. An intralaboratory blind sample program is performed on selected samples. The blinds are prepared from samples sent from a cross check laboratory and split between several analysts as determined by the Chemistry Operations Supervisor or designee. This process allows evaluation of individual analysts performance while at the same time satisfying the blind duplicate and cross check laboratory requirements.

Section C.11.c (Calibrations) suggests that appropriate standards be used to calibrate

continuous radioactivity monitors and that the relationship be established between monitor readings and concentration over the full range of the readout device. In those cases where mixed fission gases or corrosion and activation products are not available, vendor instrument performance data or calculations will be used. Subsequent inservice calibrations will be performed using the specific radionuclide analytical results from grab samples taken from the effluent release path.

Appendix A, Section A.3.a (1) and Section A.3.a (3), analytical frequencies are not consistent with standard sampling and analytical techniques. Improved sensitivities and C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-101 more realistic quantity measurements can be made by performing 140 Ba-La, 89-90 Sr, and gross alpha measurements on a monthly composite sample of weekly samples.

Exception is taken to the Appendix A, Section B.1.c, requirement for a special sample and analysis of one liquid waste batch per month for entrained fission and activation gases. The gamma spectrum analysis performed prior to the release of any waste liquid batch will identify such gases without performing a separate or special analysis.

The sensitivity slated in Appendix A, Section B.3, for gamma-emitting radionuclides (5 x 10-7 µCi/ml) will be applied in the case of principal gamma-emitting nuclides.

Specific Evaluation Reference

See Section 11.5.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-102 Regulatory Guide 1.22, Revision 0, February 1972 Periodic Testing of Protection System Actuation Functions

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The CGS protection system and the systems whose operation it initiates are designed to permit periodic testing of the actuation devices during reactor operation. The periodic tests will duplicate, as closely as practical, the performance that is required of the actuation devices in the event of an accident. The tests will be performed in overlapping portions so that an actual reactor scram will not occur as a result of the

testing.

Specific Evaluation Reference

See Section 7.3.2.1.3.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-103 Regulatory Guide 1.23, Revision 0, February 1972 Onsite Meteorological Program

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

Where conflicts exist between the recommendations specified in Regulatory Guide 1.23, Revision 0 and those recommended in Regulatory Guide 1.97, Revision 2, the Columbia Generating Station will comply with the recommendations of Regulatory Guide 1.97, Revision 2 unless noted in the text discussions as meeting Regulatory Guide 1.97, Revision 3 requirements (see Section 7.5.2.2.3).

General Compliance or Alternate Approach Assessment

The requirements of this regulatory guide for a meteorological program to provide the meteorological data required to estimate potential radiation doses to the public have

been and are being implemented for CGS.

Specific Evaluation Reference

See Sections 2.3.2 , 2.3.3 , 7.7.1 , and the Emergency Plan.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-15-032 1.8-104 Regulatory Guide 1.26, Revision 3, February 1976

Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste

Containing Components of Nuclear Power Plants

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

The definition of quality group classifications for CGS was provided in the PSAR in accordance with ASME B&PV Code, Sections III and VIII. Quality group

classifications have been maintained during design and construction. Quality group classifications are maintained during plant operations and modifications by plant administrative procedures and the plant modification control process. The quality group classifications are commensurate with the safety functions performed by the

safety-related components.

The turbine stop valves and bypass valve, which are classified Quality Group D, are subject to an enhanced quality assurance program comparable to that of Quality

Group B.

Specific Evaluation

Reference:

See Section 3.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-105 Regulatory Guide 1.27, Revision 2, January 1976 Ultimate Heat Sink for Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

Energy Northwest complies with Regulatory Guide 1.27, Revision 2, without any exceptions and with one clarification.

The clarification is that the tower makeup system (TMU) water supply is only an ultimate heat sink feature in the event of a design basis tornado. Since Regulatory

Guide 1.27 states that we need not consider two or more most severe natural phenomena occurring simultaneously, the TMU was designed to be tornado proof but was not designed and constructed to withstand the effects of the operating basis earthquake (OBE) and water flow based on severe historical events in the region.

Specific Assessment Reference

See Section 9.2.5.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-106 Regulatory Guide 1.28, Revision 0, June 1972 Quality Assurance Program Requirements (Design and Construction)

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide as described below.

General Compliance or Alternate Approach Assessment

Procurement documents issued after November 1973 required compliance with ANSI N45.2. Prior to that time, an explanative version of 10 CFR 50 Appendix B was used. The design and construction activities initially complied with 10 CFR 50 Appendix B. In November 1974, reference to ANSI N45.2 was added to the construction specifications.

ANSI N45.2 does not apply to the activities covered by Section III and Section XI of the ASME Code; however, the quality assurance program requirements may be extended to

these activities based on project requirements.

Specific Evaluation Reference

None C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-107 Regulatory Guide 1.29, Revision 3, September 1978 Seismic Design Classification

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

CGS classifications are consistent with Regulatory Guide 1.29 with the following clarification:

Cooling of the spent fuel storage pool is accomplished by the spent fuel cooling and cleanup system or by the seismic category RHR cross connection. The spent fuel pool cooling portion which is used normally to cool the spent fuel pool water was Seismic

Category I by the first refueling outage. The cleanup portion of the system is not Seismic Category I. However, all structures, systems, and components required for maintaining water cover for the spent fuel are Seismic Category I. The spent fuel

cooling system uses some common pump suction and discharge piping which is

embedded in concrete. Prior to the first refueling outage, the Seismic Category I RHR

system cross connection would have been used in case of core offload (see

Section 9.1.3).

Specific Evaluation Reference

See Sections 3.2.1 , 3.7 , 3.8 , 3.9 , 3.10 , 9.1.3 , and the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-108 Regulatory Guide 1.30, Revision 0, August 1972 Quality Assurance Requirements for the Installation, Inspection, and Testing of

Instrumentation and Electrical Equipment.

I Design and Construction Phase

Compliance or Alternate Approach Statement

CGS generally complies with the guidance set forth in this regulatory guide. In a few cases, CGS complied with the intent of this guidance by an alternate approach.

General Compliance or Alternate Approach Assessment

Procurement documents require compliance with ANSI N45.2.4 for the installation, inspection, and testing activities performed, except in those isolated instances where requirements were entered directly in the specification with limited or no reference to

ANSI N45.2.4 or IEEE 336.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-109 Regulatory Guide 1.31, Revision 3, April 1978 Control of Ferrite Content in Stainless Steel Weld Metal

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

CGS complies fully with Revision 3 of this guide on all contracts initiated after the date

of its publication. Prior to issuance of Revision 3, CGS conformed to Revision 2 of this regulatory guide.

Specific Evaluation Reference

See Sections 4.5.2.4 , 5.2.3.3 , and 5.3.1.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-110 Regulatory Guide 1.32, Revision 2, February 1977 Criteria for Safety Related Electric Power Systems for Nuclear Power Plants

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in Revision 0 of this regulatory guide.

(Revisions 1 and 2 are not applicable to CGS since they are for use in evaluations of construction permits docketed after November 1, 1976, and April 15, 1977, respectively.)

General Compliance or Alternate Approach Assessment

The CGS design is in full compliance with both Revision 0 of this regulatory guide and with Revision 2 of this regulatory guide, with the exception of those sections of the regulatory guide which require compliance with Regulatory Guides 1.93, Revision 0, and 1.75, Revision 0. See Section 8.3.1.2.1.1 for analysis of the CGS design relative to Regulatory Guide 1.75, Revision 0.

Specific Evaluation References

See Sections 8.1.5.1 , 8.1.5.2 , 8.2.2.4 , 8.3.1.1.7.1 , 8.3.1.2.1.3 , 8.3.1.3 , 8.3.1.4 , 8.3.2.1.1 , 8.3.2.2.1 , 8.3.2.3 and 8.3.2.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-111 Regulatory Guide 1.33, Revision 2, February 1978 Quality Assurance Program Requirements (Operation)

Compliance or Alternate Approach Statement:

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

Compliance or Alternate Approach Assessment

Compliance is discussed in the OQAPD.

Specific Evaluation Reference

See Section 13.5.1.1 and the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-112 Regulatory Guide 1.34, Revision 0, December 1972 Control of Electroslag Weld Properties

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since electroslag welding has not been used for welding of Class 1 or 2 vessels or components fabricated of low alloy or

austenitic steel.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-113 Regulatory Guide 1.35, Revision 2, January 1976 Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since CGS does not have a prestressed concrete containment structure with ungrouted tendons.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-114 Regulatory Guide 1.36, Revision 0, February 1973 Nonmetallic Thermal Insulation for Austenitic Stainless Steel

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

Thermal insulation on stainless steel piping conforms to requirements of this regulatory

guide.

Specific Evaluation

Reference:

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-115 Regulatory Guide 1.37, Revision 0, March 1973 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of

Water-Cooled Nuclear Power Plants

I Design and Construction Phase

Compliance or Alternate Statement

CGS generally complies with the guidance set forth in this regulatory guide. In a few cases, CGS complied with the intent of this guidance by an alternate approach.

General Compliance or Alternate Approach Assessment

Procurement documents generally required compliance with ANSI N45.2.1. Whether or not reference to ANSI N45.2.1 was provided, a detailed specification section supplied comprehensive instructions on cleaning and cleanliness.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-116 Regulatory Guide 1.38, Revision 2, May 1977 Quality Assurance Requirement for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants

I Design and Construction Phase

Compliance or Alternate Approach Statement

CGS generally complies with the guidance set forth in Revision 0 of this regulatory guide. In a few cases, CGS complied with the intent of this guidance by an alternate

approach.

The changes to the regulatory positions of Revision 1 and 2 of this regulatory guide, which specify additional detailed requirements and make certain nonmandatory sections

of ANSI N45.2.2 mandatory, are not implemented.

General Compliance or Alternate Approach Assessment

Procurement documents required compliance with ANSI N45.2.2, Revision 0, and/or contained a generic specification packaging section and/or specified directly

requirements for these functions.

The regulatory positions contained in Revision 1 and 2 of this regulatory guide changed significantly from the original issue. Revision 1 and 2 contain additional detailed requirements and make nonmandatory sections of ANSI N45.2.2 mandatory. Some, but not all, of the changes to the regulatory positions are included in procurement documents. Since these changes were made after award of the applicable procurement documents, Revision 1 and 2 are not fully implemented.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-117 Regulatory Guide 1.39, Revision 1, October 1976 Housekeeping Requirements for Water-Cooled Nuclear Power Plants

I Design and Construction Phase

Compliance or Alternate Approach Statement

CGS generally complies with the guidance set forth in this regulatory guide. In some cases, CGS complied with the intent of this guidance by an alternate approach.

General Compliance or Alternate Approach Assessment

Procurement documents required compliance with ANSI N45.2.3 or with selected

portions of ANSI N45.2.3 or specified directly applicable housekeeping requirements.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-118 Regulatory Guide 1.40, Revision 0, March 1973 Qualification Tests of Continuous Duty Motors Installed Inside the Containment of

Water-Cooled Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in the regulatory guide.

General Compliance or Alternate Approach Assessment

Containment fans have been qualified for in containment use in accordance with IEEE 334-1974.

Specific Evaluation Reference

See Section 9.4.11.3.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-119 Regulatory Guide 1.41, Revision 0, March 1973 Preoperational Testing of Redundant On-Site Electrical Power Systems to Verify Proper Load

Group Assignments

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in the regulatory guide.

General Compliance or Alternate Approach Assessment:

As part of the preoperational test program, the onsite electric power systems will be tested in order to verify the existence of independence among redundant onsite power

sources and their respective load groups.

Specific Evaluation Reference

See Sections 8.1.5.2 , 8.3.1.2.2 and 14.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-120 Regulatory Guide 1.43, Revision 0, May 1973 Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since CGS does not use stainless steel cladding on coarse grain low-alloy steel for safety class components.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-121 Regulatory Guide 1.44, Revision 0, May 1973 Control of the Use of Sensitized Stainless Steel

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

CGS conforms fully to the recommended welding controls for stainless steel welding.

All materials are purchased to the latest ASME and ASTM specifications at time of order, and the cleaning requirements set forth in the guide are implemented during

document review of vendor cleaning procedures.

Specific Evaluation Reference

See Sections 4.5.2.4 and 5.3.1.4.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-005 1.8-122 Regulatory Guide 1.46, Revision 0, May 1973

Protection Against Pipe Whip Inside Containment

Compliance or Alternate Approach Statement:

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

Pipe break location criteria is based on guidelines provided in this regulatory guide, as

well as the NRC Branch Technical Positions ASB 3-1, Appendix B , and MEB 3-1. The criteria is applicable to all piping systems inside as well as outside containment.

Pipe whip protection for the recirculation system is provided by the NSSS supplier.

Pipe whip protection for all other piping systems, including the NSSS-furnished main

steam piping, is provided by the architect-engineer.

Specific Evaluation

Reference:

See Section 3.6.2.1.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-15-008 1.8-123 Regulatory Guide 1.47, Revision 0, May 1973

Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems

Compliance or Alternate Approach Statement:

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment:

The alternate approach is provided in Section 7.1.2.4.

Specific Evaluation

Reference:

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-15-008 1.8-124

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C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 1.8-125 Regulatory Guide 1.48, Revision 0, May 1973 Design Limits and Loading Combinations for Seismic Category I Fluid System Components

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

Implementation of this regulatory guide is discussed in Section 3.9.3.1.1.7.

Specific Evaluation

Reference:

See Section 3.9.3.1.1.7.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 1.8-126 Regulatory Guide 1.50, Revision 0, May 1973 Control of Preheat Temperature for Welding Low-Alloy Steel

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

CGS complies with the guidance set forth in the regulatory guide by maintaining the preheat temperature of low alloy steel welds until the post-weld heat treatment has been performed. For welds which were made without this keep hot requirement, Regulatory Position C4 for determining the soundness of the weld by acceptable examination procedures, has been enforced.

Specific Evaluation

Reference:

See Section 5.3.1.4.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 1.8-127 Regulatory Guide 1.51, Revision 0, May 1973 In-Service Inspection of ASME Code Class 2 and 3 Nuclear Power Plant Components

Compliance or Alternate Approach Statement:

This regulatory guide has been withdrawn and is no longer applicable.

General Compliance or Alternate Approach Assessment:

Inservice inspection of CGS is based on ASME Section XI for Classes 1, 2, and 3.

Specific Evaluation

Reference:

See Section 3.9.6.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-011 1.8-128 Regulatory Guide 1.52, Revision 2, March 1978

Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants

Compliance or Alternate Approach Statement:

CGS complies with the intent of the guidance given in Revision 2 of this regulatory guide with the exception that the operation identified in C-4.d, Revision 2 is modified by Technical Specification Amendment 239 which adopts the testing criteria identified in Regulatory Position 6.1 of Revision 3.

General Compliance or Alternate Approach Assessment:

Standby gas treatment filter units and the control room emergency filter units are required to perform safety-related functions. A comparison of the engineered safety feature air filtration systems with respect to the regulatory position of Regulatory Guide 1.52, Revision 2, Article C, is as follows:

Paragraph Number SGTS Control Room System

C-1. Environmental Design Criteria

1.a In compliance In compliance 1.b In compliance In compliance 1.c In compliance In compliance 1.d In compliance In compliance 1.e In compliance In compliance

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 1.8-129 C-2. System Design Criteria

2.a In compliance See Note 1 2.b In compliance In compliance 2.c In compliance In compliance 2.d See Note 2 See Note 2 2.e In compliance In compliance 2.f In compliance In compliance 2.g See Note 3 See Note 3 2.h In compliance In compliance 2.i In compliance In compliance 2.j See Note 4 See Note 4 2.k In compliance In compliance 2.1 In compliance In compliance C-3. Component Design Criteria and Qualification Testing 3.a See Note 5 See Note 5 3.b In compliance In compliance 3.c In compliance In compliance 3.d See Note 6 See Note 6 3.e In compliance In compliance 3.f In compliance In compliance 3.g See Note 7 See Note 7 3.h In compliance In compliance 3.i See Note 8 See Note 8 3.j In compliance In compliance 3.k In compliance In compliance 3.l In compliance In compliance 3.m In compliance In compliance 3.n In compliance In compliance 3.o In compliance In compliance 3.p In compliance In compliance

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 1.8-130 C-4. Maintenance 4.a See Note 9 See Note 9 4.b See Note 10 See Note 10 4.c In compliance In compliance 4.d See Note 11 In compliance 4.e In compliance In compliance C-5. In-Place Testing Criteria 5.a In compliance In compliance 5.b See Note 13 In compliance 5.c See Note 14 See Note 14 5.d See Note 14 See Note 14 C-6. Laboratory Testing Criteria For Activated Carbon 6.a See Note 12 See Note 12 6.b See Note 12 See Note 12 Note 1 (C-2.a) Demisters are not provided in the control room filter units due to the absence of entrained moisture during normal and abnormal conditions. High-efficiency particulate air (HEPA) filters are not

provided after the charcoal filter because filter unit discharges

into control room air conditioning unit on intake side of medium

efficiency filters.

Note 2 (C-2.d) Both units of the standby gas treatment system are located in secondary containment and are not subject to containment pressure surges during accidents. Redundant Seismic Category I

valves in series isolate and protect these units from containment

DBA pressures. Both units of the control room emergency filter system are not subject to containment pressure surges during

accidents.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 1.8-131 Note 3 (C-2.g) Abnormal pressure drops across critical components of the SGTS and control room filter units cause an alarm in the main control room, however, no facilities to record the pressure drops are provided. A record of pressure drop across individual

components and the total SGTS system would be of no value because the SGTS is a variable flow system, with flow modulated to maintain the reactor building at a fixed negative pressure.

Flow through the system, which is the pertinent parameter, is recorded in the main control room, and computer input is provided to record high pressure alarms across critical components.

Note 4 (C-2.j) SGTS filter units are not designed to be removable from the building as an intact unit. The size of the units precludes removal

in one section. In the event the units become radioactively contaminated they will be permitted to decay in place until radiation levels are sufficiently low to permit the removal of all

internals for disposal.

Note 5 (C-3.a) SGTS system demisters furnished by FARR Company, are not in complete conformance with ANSI N509-1976 because they were

not qualified by testing in accordance with AEC report MSAR-71-45. A moisture eliminator study performed by FARR Company in 1970, which did not conform to the MSAR-71-45 test setup, indicated that the installed demisters will protect the HEPA filters in the system from blinding under conditions far

more severe than those hypothesized for the SGTS system.

Since, under the accident mode, entrained water droplets will not be in the inlet air stream, the FARR tests and qualification are

considered adequate.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-011 1.8-132 Note 6 (C-3.d) HEPA filters are not subjected to iodine removal sprays, therefore, aluminum separators are used.

An alternate approach to determine acceptable design and qualification testing of HEPA filters is the use of Regulatory Guide 1.52, Revision 3, Section 4.4.

Note 7 (C-3.g) Access doors into SGTS units are 50 x 20 in. Vacuum breakers are not provided on doors of SGTS and control room units. Unit

fans are normally off.

Note 8 (C-3.i) Test 4, Activity (Ref. Table 5-1, ANSI N509-1976)

Base carbon (unimpregnated) activity test was not previously required. Because all available carbon was of the impregnated type this was not run.

Test 5, Radioiodine Removal Efficiency (Ref. Table 5-1, ANSI N509-1976)

New carbon will be tested in accordance with ASTM D3803-1989.

Average atmosphere resident time in each SGTS unit is greater

than 0.5 sec.

Note 9 (C-4.a) Doors provided on SGTS Units are 50 x 20 in. Access panels are provided on control room units. Vacuum breakers are not provided on any of the units since they are normally not

operational.

Note 10 (C-4.b) Control room filter units have approximately 18 in. between prefilter and HEPA filter frames, and approximately 4 ft are provided between HEPA and charcoal filter frames. SGTS filter units have a minimum of three feet provided between demister, heater, prefilter, HEPA and charcoal filter frames.

Note 11 (C-4.d) Strip heaters are provided in the charcoal filter plenum of the SGTS units to maintain charcoal beds moisture free, therefore, operation of the fans is not required for that purpose. Testing of the blast coil heaters is performed in accordance with the requirements of Regulatory Position 6.1 of Revision 3 of the Regulatory Guide 1.52 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-133 Note 12 The laboratory testing criteria for the carbon adsorber section (C-6.a C-6.b) of the SGTS and CREF System meets the objectives of this section of the guide. Twelve representative test samples of four-inch length are provided across each of the two 4 in. deep beds in each SGTS filter unit. At least once per 30 months one sample from across each SGT and CREF adsorber bed is removed and sent to a laboratory for testing. For the SGTS, samples are tested in series to represent the 8-inch total bed depth. Laboratory tests are performed in accordance with ASTM D3803-1989 with methyl iodide at 30°C and 70% relative humidity with a penetration of less than 0.5% for the SGTS and less than 2.5% for the CREF System as an acceptance level. The SGTS will also be tested at a face velocity of 75 ft per minute. In the event that a sample fails this test, the carbon adsorber in its bed will be replaced.

Note 13 (C-5.b) The flow distribution tests developed by the designer combined with the series filter design at CGS adequately meet the intent of this test. The results of the flow distribution tests as set forth in ANSI N51 are difficult to interpret with the U shaped charcoal

beds installed due to air flow disturbance caused by the

measuring apparatus. This is particularly true on the parallel legs of the U shaped beds, where the flow measuring device must be placed in the rather narrow air passage. Flow distribution criteria

was developed by the designers based on the +/-20% variation criteria established in Regulatory Guide 1.52 and has been met in

field tests. In addition, each of the filter trains has two separate

charcoal beds in series. This allows mixing of the filtered gas

between the beds and further reduces the effects of variations in

charcoal packing distribution.

Note 14 The inplace leak testing of the SGT and CREF HEPA and carbon (C-5.c C-5.d) filters meets the objectives of this section of the guide with the exception that testing is performed in accordance with

ASME N510-1989, Sections 10 and 11, respectively.

Specific Evaluation Reference

See Section 6.5.1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-134 Regulatory Guide 1.53, Revision 0, June 1973 Application of the Single-Failure Criterion to Nuclear Power Plant Protective Systems

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in the regulatory guide.

General Compliance or Alternate Approach Assessment

Regulatory Guide 1.53 provides guidance for the application of the single-failure

criterion as discussed in IEEE 379-1972. The regulatory guide recommends the application of IEEE 379-1972 with four supplemental conditions. The design of the CGS electrical system is in conformance with IEEE 379-1972 and the four supplemental conditions noted in Regulatory Position C.

Specific Evaluation Reference

See Section 8.1.5.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-135 Regulatory Guide 1.54, Revision 0, June 1973 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear

Power Plants

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide as described below.

General Compliance or Alternate Approach Assessment

Special decontaminable coatings in primary containment areas are manufactured and applied in accordance with quality assurance requirements of ANSI N101.4.

Specific Evaluation Reference

See Section 6.1.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-136 Regulatory Guide 1.55, Revision 0, June 1973 Concrete Placement in Category I Structures

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The requirements of the guide have been included in the appropriate construction contract specifications. Compliance with the guide is assured by the application of appropriate concrete specifications, construction practices, codes and standards, including the documents recommended by the guide, for the placement of concrete; by the implementation of approved communications procedures between qualified design

and construction forces; and by implementation of an approved QA program which ensures design control and coordinated quality control of concrete material, placement, inspection and testing between applicant, designer and constructor.

Specific Evaluation Reference

See Sections 3.8.3.2 , 3.8.3.6 , 3.8.4.2 , 3.8.4.6 , and 3.8.5.2 and Table 3.8-4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-137 Regulatory Guide 1.56, Revision 0, June 1973 Maintenance of Water Purity in Boiling Water Reactors

I. Design and Construction Phase

Compliance or Alternate Approach Statement

The design of CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

CGS design complies with the guidance of this regulatory guide by providing for the

following:

a. Conductivity measurement and recording of the condenser hotwell and condensate flow discharge to the condensate demineralizer system,

b. Flow measurement and recording of flow through each condensate demineralizer unit,

c. Conductivity measurement, recording, and alarming of the condensate effluent discharge from each condensate demineralizer unit and from the combined

system effluent,

d. Conductivity measurement, recording, and alarming of the inlet and outlet coolant to and from the RWCU system,

e. Extensive sampling of reactor coolant and auxiliary systems,

f. Full flow condensate demineralizer system, and

g. Excess condensate demineralizer capacity to permit recharging of resin beds during normal plant operation.

Specific Evaluation Reference

See Section 5.2.3.2.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-138 II. Operations Phase Compliance or Alternate Approach Statement

Operation of CGS RWCU and condensate demineralizer system complies with the general guidance set forth in Revision 1, July 1978, of this regulatory guide.

General Approach or Alternate Approach Assessment

Operation of CGS complies with the guidance of the regulatory guide by providing the

following:

a. Operating limits are prescribed for condensate filter demineralizers. Plant operating conductivity limits are defined for the RWCU demineralizers.

Effluent conductivity for the individual demineralizers is recorded and a main control room alarm is triggered when conductivity limits are reached or exceeded;

b. Condensate filter demineralizer conductivity and flow instrumentation are used in the general assessment of individual demineralizer unit performance and

capacity;

c. An operational limit is set for hotwell conductivity which triggers a main control room alarm. Hotwell conductivity, in conjunction with precalculated assessment

of condenser inleakage rates and demineralizer performance permits appropriate

action to be taken on exceeding the operating limit setpoint;

d. Laboratory analyses are performed for chloride, pH, and conductivity at intervals appropriate to the plant operating status. Sampling and analysis

frequency is described in the LCS and plant procedures; and

e. Not applicable exception is taken to item C.4.d which applies to bead-type, deep-bed demineralizer systems, which are not incorporated into the CGS design. The general guidance of this item will, however, be applied to the pressure precoat filter demineralizer systems. Each lot of precoat resins will be analyzed for capacity and impurity levels. Frequency of precoat changeout will be staggered and is initially dictated by pressure drop associated with suspended

solids. Subsequent to pressure drop limitations, frequency of sequential precoat changeout is established based on dissolved chemical constituents and flow

throughput parameters.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-139 Regulatory Guide 1.57, Revision 0, June 1973 Design Limits and Loading Combinations for Metal Primary Reactor Containment System

Components

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

The structural design criteria for the primary containment vessel is consistent with the provisions of this regulatory guide, except with respect to the stress limits specified in Section C-1-b(2) of the guide, for the load combination of accident recovery flooding plus OBE. For this load combination, the stress limits used for CGS are within the limits set forth in the NRC Standard Review Plan Section 3.8.2, Table 3.8.2-1.

This exception has precedent as stated in GESSAR, paragraph 3.8.2.3.12, Accident Recovery Evaluation, Page 3.8-9b, and has been accepted by the NRC, as documented in paragraph 3.8.2, page 3-14, of the NRC Safety Evaluation Report for the GESSAR-328 Nuclear Island Standard Design dated December 1975.

Specific Evaluation Reference

See Section 3.8.2.3.10.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-140 Regulatory Guide 1.58, Revision 1, August 1980 Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel

I Design and Construction Phase

Compliance or Alternate Approach Statement

As of November 1980, CGS complies with the guidance set forth in this regulatory guide via an alternate approach described below.

General Compliance or Alternate Approach Assessment

Prior to issuance of Revision 1 of this Regulatory Guide, personnel performing quality-related activities were provided indoctrination and training in the requirements of the applicable quality assurance program, procedures, instructions and drawings affecting their work. Documented evidence of the above training was maintained. The

indoctrination and training complied with the requirements of Appendix B , 10 CFR Part 50, and ANSI N45.2.

As of November 1980, in addition to the indoctrination and training requirements noted above, requirements which meet this regulatory guide were imposed on site contractors for personnel performing inspections, examinations, and tests. These requirements specify that initial evaluations of education, experience, and qualifications are to be performed and documented; however, formal certificates are not required to be issued

because specific inspections, examinations, and tests are performed in accordance with approved procedures. Therefore, specific capability identification and levels of

certification are not required.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD. Also see Section 14.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-141 Regulatory Guide 1.59, Revision 1, April 1976 Design Basis Floods for Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

All the requirements that are specified in Regulatory Guide 1.59 are followed in the

design of CGS.

Based on Regulatory Guide 1.102, the plant site is classified as Dry Site. Therefore, CGS is considered to be in compliance with Regulatory Guide 1.59 and its

Appendix A.

Specific Evaluation Reference

See Section 2.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-142 Regulatory Guide 1.60, Revision 1, December 1973 Design Response Spectra for Seismic Design of Nuclear Power Plants

Compliance or Alternate Approach Statements

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

CGS meets the seismic requirements previously acceptable to the NRC as discussed in Section 3.7.1.1.

Specific Evaluation Reference

See Section 3.7.1.1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-143 Regulatory Guide 1.61, Revision 0, October 1973 Damping Values for Seismic Design of Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The damping values recommended by Regulatory Guide 1.61 are greater, and therefore less conservative, than the values used for CGS. The more conservative CGS design satisfies the requirements of Regulatory Guide 1.61.

Specific Evaluation Reference

See Section 3.7.1.3.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-144 Regulatory Guide 1.62, Revision 0, October 1973 Manual Initiation of Protective Actions

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

Means are provided in the main control room for the manual initiation of BOP engineered safety feature systems or supporting systems at the division level by the

operation of a minimum of equipment.

Specific Evaluation Reference

See Section 7.3.2.1.3.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-145 Regulatory Guide 1.63, Revision 2, July 1978, and Revision 3, February 1987 Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear

Power Plants

Compliance or Alternate Approach Statement

Revisions 2 and 3 are not applicable to CGS since they apply to the evaluation of construction permit applications docketed after August 31, 1978 and February 28, 1987, respectively. CGS complies with the guidance set forth in IEEE 317-1972 as modified by Revision 0 of Regulatory Guide 1.63.

General Compliance or Alternate Approach Assessment

The compliance assessment given below correspond numerically to the regulatory positions as indicated in Section C of Regulatory Guide 1.63, Revision 0, October 1973.

1. Capability of withstanding maximum fault I 2T heating in the case that overload protective devices fail:

CGS is in compliance with this requirement. In all cases, the overcurrent protective devices in circuits subject to short circuit are backed up by other overcurrent protective devices which are also designed to limit the fault current

I 2T heating experienced by the penetration conductors to levels below the conductor ratings.

2. The maximum containment pressure specified for CGS complies with the safety margins required by the ASME B&PV Code, Article N3000, footnote 1.

3. The position refers to specific applicability or acceptability of other codes, standards, and guides covered separately in other regulatory guides.

4. CGS complies with the requirement of IEEE 336 and ANSI N45.2 concerning the QA.

Specific Evaluation

Reference:

See Sections 3.8.6 , 7.1.2.3 , and 8.1.5.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-146 Regulatory Guide 1.64, Revision 2, June 1976 Quality Assurance Requirements for the Design of Nuclear Power Plants

I. Design and Construction Phase

Compliance or Alternate Approach Assessment

Regulatory Guide 1.64, Revision 0, Revision 1, and Revision 2 do not apply to CGS since they apply to construction permits docketed after September 1973.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

II. Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-147 Regulatory Guide 1.67, Revision 0, October 1973 Installation of Overpressure Protection Devices

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since the reactor coolant system pressure boundary safety/relief valve relieves to a closed discharge system.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-148 Regulatory Guide 1.68, Revision 1, January 1977 Initial Test Programs for Water-Cooled Reactor Power Plants

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to the CGS initial test program since Revision 0

of this regulatory guide is committed to in Section 14.2.7. However, CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

See Section 14.2 for description of initial testing program and to Sections 14.2.7 and 1.8.2 for statements concerning compliance with Regulatory Guide 1.68, Revision 0. Revision 1 of this guide in general clarifies Revision 0 and therefore there are no

exceptions to the intent of this procedure.

Specific Evaluation Reference

See Sections 14.2.7 and 1.8.2 for a discussion of Regulatory Guide 1.68, Revision 0.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-149 Regulatory Guide 1.68.1, Revision 1, January 1977 Preoperational and Initial Startup of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants.

Compliance or Alternate Approach Statements

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessments:

The preoperational testing and the initial Startup testing as described in Section 14.2 complies with the intent of this regulatory guide. However, due to the limitations of the auxiliary steam supply system, the confirmation that the feedwater pumps satisfy

required head, flow rate and suction head will not occur until the startup phase of the initial test program when the normal steam supply is available to the feedwater pump

turbines.

Specific Evaluation Reference

See Section 14.2.12.1.1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-150 Regulatory Guide 1.68.2, Revision 0, January 1977 Initial Startup Test Program To Demonstrate Remote Shutdown Capability For Water-Cooled Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate approach assessment

The startup test described in Section 14.2.12.3.28 complies with the regulatory guide with the following exceptions:

a. The test will be initiated by scramming plant from the control room versus a location outside the control room as described in Section C.3 of the regulatory guide. This exception is made to better simulate the actual procedure which

would be followed if a control evacuation were to occur. The capability to scram the reactor outside the control room exists; for example, tripping the RPS

motor generator (MG) sets.

b. The cold shutdown demonstration procedure as described in Section C.4 of the Regulatory Guide may not be performed immediately following the

demonstration of achieving and maintaining safe hot standby from outside the

control room. Rather this cooldown portion may be performed when cooldown

is required during the course of the normal power ascension test program.

Although this is an exception to Regulatory Guide 1.68.2, Revision 0, Revision 1 of this Guide contains provisions for a delay in the demonstration of

cooldown.

Specific Evaluation Reference

See Sections 14.2.12.3.28 and 7.4.1.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-151 Regulatory Guide 1.69, Revision 0, December 1973 Concrete Radiation Shields for Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

Although the regulatory guide was promulgated after design and specification implementation of the engineering criteria, the recommended design and construction practices specified in the regulatory guide are documented in codes and specifications

which were used in the development of the engineering criteria and contract specifications.

Specific Evaluation Reference

See Section 12.3.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-152 Regulatory Guide 1.70, Revision 2, September 1975 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants - LWR

Edition

Compliance or Alternate Approach Statement

This FSAR complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The FSAR has generally been prepared to satisfy the requirements of Regulatory Guide 1.70, Revision 2. This includes both format and content.

Specific Evaluation Reference

The balance-of-plant (BOP) portions of this FSAR.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-153 Regulatory Guide 1.71, Revision 0, December 1973 Welder Qualifications for Areas of Limited Accessibility

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

There are few incidents where welding accessibility is limited during fabrication. Where accessibility to any weld joint was restricted to a degree which prevented the welder from direct visual observation of the arc and the puddle in any area of the weld, or which required the use of mirrors or extensions to the torch handle or electrode holder, the contractor notifies the welding engineer. All limited access welds are determined by a welding engineer. For ASME Section III, Class 1, 2, and 3 components and Subsection NF and NE, a performance qualification test that simulates the limited access condition is required by the welding engineer. For welds in the pressure retaining components the welders test weld is radiographed in accordance with and shall conform to the acceptance standards of ASME Section VIII, Division 1, U.W.-51. Alternately, the weld may be examined ultrasonically in accordance with

ASME Section VIII, Division 1, Appendix U.

Specific Evaluation Reference

See Sections 4.5.2.4 , 5.2.3.3 , and 5.3.1.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-154 Regulatory Guide 1.72, Revision 0, December 1973 Spray Pond Plastic Piping

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS because CGS does not use plastic piping

in its spray ponds.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-155 Regulatory Guide 1.73, Revision 0, January 1974 Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

Auxiliary equipment associated with valve operators are tested in accordance with the subject standards. Designed service conditions are implemented in the tests.

Conservative values of the environmental variables during and after a design basis accident are used in the tests to assure that the testing is carried out under more severe environmental conditions than those expected.

Specific Evaluation Reference

See Sections 3.11 and 8.1.5.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-156 Regulatory Guide 1.74, Revision 0, February 1974 Quality Assurance Terms and Definitions

I Design and Construction Phase

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The terms used in describing and implementing quality assurance programs for CGS have complied with ANSI N45.2.10-1973 or were clarified at the point of application.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-157 Regulatory Guide 1.75, Revision 1, January 1975 Physical Independence of Electric Systems

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed after February 1974. However, CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

See Section 8.3.1.4.2.7 for an assessment of CGS relative to this regulatory guide.

Specific Evaluation Reference

See Section 8.3.1.4.2.7.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-158 Regulatory Guide 1.76, Revision 0, April 1974 Design Basis Tornado for Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

The tornado design criteria for Columbia Generating Station were revised based on design basis tornado characteristics in NUREG-1503. The design basis tornado characteristics used are less severe than those specified in Regulatory Guide 1.76 for

Region III. In January 1996, the revised criteria were found acceptable by the NRC.

Specific Evaluation Reference

See Section 3.3.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-159 Regulatory Guide 1.78, Revision 0, June 1974 Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The main control room habitability during a postulated hazardous chemical release evaluation complies with assumptions and toxicity limits in Revision 0 of this regulatory

guide. The evaluation uses toxicity limits presented in Revision 1 for those chemicals not discussed in Revision 0. The results are presented in Chapter 6.

Specific Evaluation Reference

See Sections 2.2.3 and 6.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-160 Regulatory Guide 1.80, Revision 0, June 1974 Preoperational Testing of Instrument Air Systems

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The primary containment instrument air system preoperational test procedure

incorporated the requirements of this regulatory guide.

Specific Evaluation Reference

See Sections 14.2.7.3 and 14.2.12.1.34.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-161 Regulatory Guide 1.82, Revision 0, June 1974 Sumps for Emergency Core Cooling and Containment Spray Systems

Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to CGS since no sumps are used for ECCS and

containment spray.

General Compliance or Alternate Approach Assessment:

Not applicable.

Specific Evaluation

Reference:

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-004 1.8-162 Regulatory Guide 1.84

Design, Fabrication, and Materials Code Case Acceptability, ASME Section III Regulatory Guide Intent

This guide lists all Section III Code Cases that the NRC has approved for use. It is updated on a regular basis to reflect the changes to the ASME Code Cases and the current position of the NRC on acceptability for use. The guide contains tables that detail the NRC acceptance requirements for current, annulled, and superseded Code Cases. Code Cases that the NRC determined to be unacceptable are listed in Regulatory Guide 1.193, ASME Code Cases Not Approved for Use.

Application Assessment

Assessed capability in design.

Compliance or Alternate Approach Statement

The current version of the Regulatory Guide is utilized to determine acceptable Code Cases for all new and existing plant applications. The FSAR does not track individual Code Cases and revision numbers. Not all acceptable Code Cases listed in the regulatory guide are used. The Code Cases that are utilized for Columbia are referred to in the plant design/installation documentation.

General Compliance or Alternate Approach Assessment

Code Cases are utilized in accordance with the requirements of the regulatory guide provisions for acceptance.Section III Code Cases that are not yet endorsed may be utilized via submittal to the NRC for approval in accordance with the regulatory guide. The plant scope of supply is in full compliance with this regulatory guide.

Specific Evaluation Reference

See Section 3.8.2.2.

Similar Application Reference

None.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-004 1.8-163 Regulatory Guide 1.85, Revision 31, 1998

• Materials Code Case Acceptability - ASME Section III, Division 1

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide as described below.

General Compliance or Alternate Approach Assessment

The use of an ASME Section III, Division 1, code case applicable to materials use on

CGS is approved by Energy Northwest only after evaluating its technical acceptability and confirming that its use is acceptable to the NRC. This confirmation is by ascertaining that the code case is listed in this regulatory guide (or applicable earlier

revision) or by specific written acceptance by the NRC.

Specific Evaluation Reference

See Section 3.8.2.2.

• Regulatory Guide 1.85 was withdrawn in 2004. See Regulatory Guide 1.84 for NRC acceptance of current Materials Code Cases.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-164 Regulatory Guide 1.88, Revision 2 October 1976 Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records

I Design and Construction Phase Compliance or Alternate Approach Statement

I Design and Construction Phase Prior to the original issue of this regulatory guide and construction of the CGS records facility, Project Quality Assurance complied with the intent of 10 CFR Part 50, Appendix B, by duplicate storage of records. Project Quality Assurance also complied with the original issue and revisions of this regulatory guide by duplicate storage.

Since March 1977, Project Quality Assurance has complied with Revision 2 of this

regulatory guide as described below.

General Compliance or Alternate Approach Assessment:

Since March 1977, the collection, storage, and maintenance of quality assurance records by Project Quality Assurance has been in compliance with ANSI N45.2.9 and

NFPA No. 232-1975 for fire protection as imposed by this regulatory guide. The

record facility has a minimum of a 2-hr rating.

Procurement documents directly specify requirements for collection, storage, and maintenance of records. The requirements generally meet the intent of ANSI N45.2.9 except that storage facilities or cabinets are only required to meet a 1-hr rating.

II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-020 1.8-165 Regulatory Guide 1.89, Revision 1, June 1984

Qualification of Class 1E Equipment for Nuclear Power Plants

Regulatory Guide Intent:

Regulatory Guide 1.89 endorses both the requirements and recommendations of

IEEE 323-1974, IEEE Standard for Qualifying Class 1E Equipment for Nuclear

Power Generating Stations. Additional regulatory position stipulations are also included.

Compliance or Alternate Approach Statement:

CGS complies with this regulatory guide for equipment requiring environmental qualification procured after February 22, 1983.

General Compliance or Alternate Approach Statement:

For equipment requiring environmental qualification installed prior to February 22, 1983, CGS follows the guidance in NUREG-0588 Cat II.

In view of the NRC Memorandum and Order (CLI-80-21), dated May 23, 1980, all environmental qualifications of Class 1E equipment located in harsh environments are reevaluated for compliance with NUREG-0588, Category II. Where significant deviation from those guidelines is found in specific equipment qualifications, additional testing and/or analysis is performed to demonstrate the adequacy of the equipment to

perform its safety-related function.

For equipment whose qualification program has not been completed, a justification for interim operation in accordance with 10 CFR 50.49 is performed as described in the WNP-2 Environmental Qualification Report for Safety-Related Equipment, Reference 3.11-1.

Specific Evaluation

Reference:

See Section 3.11.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-166 Regulatory Guide 1.90, Revision 0, November 1974 In-Service Inspection of Prestressed Concrete Containment Structures with Grouted Tendons

Compliance or Alternate Approach Statement

This regulatory guide is not applicable because CGS does not have a prestressed

concrete containment structure with grouted tendons.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-167 Regulatory Guide 1.91, Revision 0, January 1975 Evaluation of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power

Plant Sites

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed on or after March 14, 1975. However, CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

It has been determined that the peak overpressures produced by postulated explosions occurring on transportation routes near the plant are no greater than the wind pressures

caused by the design basis tornado. Therefore, postulated explosions will not cause an accident or prevent the safe shutdown of the plant.

Specific Evaluation Reference

See Sections 2.2.1 , 2.2.2.2 , and 2.2.2.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-168 Regulatory Guide 1.92, Revision 1, February 1976 Combining Modal Responses and Spatial Components in Seismic Response Analysis

Compliance or Alternate Approach Statement:

This regulatory guide is not a requirement for CGS since it applies to the evaluation of

construction permit applications docketed after February 1976. CGS complies with the intent of the guidance set forth in this regulatory guide by implementing the regulatory guide criteria or by an alternate approach.

General Compliance or Alternate Approach Assessment:

The method of combining modal responses has been implemented in accordance with

the guides recommendations.

The combining of spatial components was performed prior to the issuance of the guide

and follows the method presented in the PSAR. The method used is an industry-accepted alternate method. The method considers the combination of the maximum structural responses to the more critical one of the two horizontal components and the vertical component of earthquake motion, using the absolute sum method. Alternatively, when the regulatory guide is followed, two horizontal components and one vertical component of earthquake motion are combined by the

square root sum of the squares method.

Specific Evaluation

Reference:

See Sections 3.7.2.6 and 3.7.2.7.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-169 Regulatory Guide 1.93, Revision 0, December 1974 Availability of Electric Power Sources

Compliance or Alternative Approach Statement

CGS complies with the regulatory position for operating the plant whenever the available electric power sources are less than the limiting conditions for operation (LCO) as defined in the regulatory guide.

General Compliance or Alternate Approach Assessment

Operating procedures incorporate the requirements of this guide.

Specific Evaluation Reference

See the Technical Specifications.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-170 Regulatory Guide 1.94, Revision 1, April 1976 Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants.

I Design and Construction Phase

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permits docketed after October 15, 1976. However, CGS complies with the intent of the guidance set forth in the guide.

General Compliance or Alternate Approach Assessment

The guidelines included in ANSI 45.2.5-1974 for installation, inspection and testing of structural concrete and structural steel, including nonpressure vessel elements of the primary containment vessel during the construction phase of CGS are reflected in the structural concrete and structural steel contract specifications for project construction.

The QA requirements of ANSI 45.2 were incorporated in these specifications.

Specific Evaluation Reference

See Sections 3.8.3.2 , 3.8.4.2 , 3.8.5.2 , and Table 3.8-4. II Operational Phase Compliance is discussed in the Topical Report referenced in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-05-009,07-025 1.8-171 Regulatory Guide 1.95, Revision 1, January 1977

Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine

Release

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since chlorine gas is not stored at CGS or nearby facilities and the expected quantities of chlorine shipped within five miles is less than the threshold volumes specified in Regulatory Guide 1.78.

Specific Evaluation Reference

See Section 6.4.4.2.

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C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-011 1.8-173 Regulatory Guide 1.100, Revision 1, August 1977

Seismic Qualification of Electric Equipment for Nuclear Power Plants

Regulatory Guide Intent

Regulatory Guide 1.100 endorses both the requirements and recommendations of IEEE 344-1975, IEEE Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, when such qualification is performed in conjunction with Regulatory Guide 1.89, and subject to the regulatory position stipulations.

Compliance or Alternate Approach Statement

This regulatory guide is applicable to CGS as clarified in Section 1.8.3 for Regulatory Guide 1.89, Revision 1 and Section 3.10.1.2.

General Compliance or Alternate Approach Assessment

All Class 1E equipment seismic qualifications are evaluated against the requirements set forth within IEEE 344-1975 as clarified in Section 3.10.1.2. The evaluations are documented and demonstrate adequacy of the methods and results of the qualifications

as equal or conservative to the requirements of IEEE 344-1975. These include

evaluations of seismic and hydrodynamic load combinations.

Specific Evaluation Reference

See Section 3.10.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-174 Regulatory Guide 1.101, Revision 1, March 1977 Emergency Planning for Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the intent set forth in this regulatory guide.

General Compliance or Alternate Approach Statement

See NUREG-0654.

Specific Evaluation Reference

See the CGS Emergency Plan.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-175 Regulatory Guide 1.102, Revision 1, September 1976 Flood Protection for Nuclear Power Plants

Compliance or Alternate Approach Statement CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The safety-related buildings and spray ponds are located far above the water level

estimated for the largest historical flood. Based on the criteria stipulated in Regulatory Guide 1.102, the CGS plant site is classified as a Dry Site.

Specific Evaluation Reference

See Section 2.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-176 Regulatory Guide 1.103, Revision 1, October 1976 Post-Tensioned Prestressed Systems for Concrete Reactor Vessels and Containments

Compliance or Alternate Approach Statement:

This regulatory guide is not applicable since CGS does not have a concrete reactor

vessel or containment.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-177 Regulatory Guide 1.104, Revision 0, February 1976 Overhead Crane Handling Systems for Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach The following safeguards are included in the design of the overhead crane:

a. Redundant low limit, main hoist,

b. Redundant equalizer bar limit switch,

c. Critical Control Path series of limit switches for the spent fuel cask handling mode, and

d. Main hoist paddle type upper limit switch to prevent the inadvertent two-blocking condition.

Specific Evaluation

Reference:

See Sections 3.8.4.1.1.5 and 9.1.4.2.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-178 Regulatory Guide 1.105, Revision 1, November 1976 Instrument Setpoints

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed after December 15, 1976.

General Compliance or Alternate Approach Assessment

Instrumentation is provided in a main control room to monitor plant variables and systems. The range of instrumentation is selected to cover the anticipated ranges of variables for the following plant conditions:

a. Normal operation,

b. Anticipated operational occurrences, and

c. Accident conditions.

To ensure adequate safety, the following plant parameters and systems are monitored and provided with appropriate controls to maintain them within

prescribed operating ranges:

1. Variables and systems that affect the fission process,

2. Variables and systems that affect the reactor core,

3. Reactor coolant pressure boundary, and

4. Containment and associated systems.

Specific Evaluation References

See Section 7.1.2.5.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-179 Regulatory Guide 1.106, Revision 1, March 1977 Thermal Overload Protection for Electric Motors on Motor Operated Valves

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed after July 15, 1976. However, CGS design complies with the intent of the guidance set forth in Section C.2 of the regulatory guide. General Compliance or Alternate Approach Assessment:

Class 1E motor-operated valve (MOV) overloads are chosen two sizes above those which would be required based on normal full load running current. The resultant

overload protection (approximately 140%) permits MOV motors to operate for extended periods at moderate overloads; tripping occurs just prior to motor damages.

Specific Evaluation Reference

See Section 8.3.1.1.9.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-180 Regulatory Guide 1.107, Revision 1, February 1977 Qualifications for Cement Grouting for Prestressing Tendons in Containment Structures

Compliance or Alternate Approach Statement This regulatory guide is not applicable to CGS because CGS does not have a

prestressed concrete containment structure with grouted tendons.

General Compliance or Alternate Approach Assessment:

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-11-002 1.8-181 Regulatory Guide 1.108, Revision 0, August 1976

Periodic Testing of Diesel Generators Used as Onsite Electric Power Systems as Nuclear

Power Plants.

Compliance or Alternate Approach Statement:

This regulatory guide is not applicable to CGS since the method described for compliance with the regulations indicated in the guide are applicable to plants having construction permit applications docketed after April 1, 1977.

General Compliance or Alternate Approach Assessment:

Preoperational and periodic testing of the diesel generators is performed as referenced

in Sections 14.2.12.1.40 and the Technical Specifications. As discussed in Section 8.3 , provisions for testability are included in the design of the standby power system.

For periodic testing, the surveillance requirements for demonstrating the operability of the diesel generators are consistent with the recommendations of Regulatory Guide 1.9 Revision 3 as described in the Bases for Technical Specification B 3.8.1. Regulatory Guide 1.9 Revision 3 includes pertinent guidance for periodic testing previously addressed in Regulatory Guide 1.108. Specific Evaluation

Reference:

See Sections 8.3 and 14.2.12.1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-182 Regulatory Guide 1.109, Revision 0, March 1976 Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents.

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide using an alternate approach.

General Compliance or Alternate Approach Statement

CGS is meeting the guidance of this regulatory guide by using Battelle Northwest models which are acceptable to the NRC.

Specific Evaluation Reference

See Sections 11.2.3.3 , 11.3.3.3 , and 5.2 of the Environmental Report.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-183 Regulatory Guide 1.110, Revision 0, March 1976 Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since a cost-benefit analysis, as described in Appendix I of 10 CFR 50 Section II-D is not required for CGS.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

See Section 11.2.3.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-184 Regulatory Guide 1.111, Revision 1, July 1977 Method for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors

Compliance or Alternate Approach Statement:

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

Analyses of atmospheric transport and dispersion of gaseous effluents at CGS are performed using the standard NRC diffusion models in NUREG/CR-2919, XOQ/DOQ:

Computer Program for the Meteorological, Evaluation of Routine Effluent Releases at Nuclear Power Stations, September 1982.

Specific Evaluation References

See Section 2.3.5.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-185 Regulatory Guide 1.112, Revision 0-R, May 1977 Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water Cooled Power Reactors.

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

The methods for calculating annual average releases of radioactive material in liquid and gaseous effluents from the plant were originally based on the GALE Code as suggested in this regulatory guide. See the sections referenced below for discussions of

the methods currently used.

Specific Evaluation Reference

See Sections 11.2.3.2 and 11.3.3.3.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-186 Regulatory Guide 1.113, Revision 1, April 1977 Estimating Aquatic Dispersion of Effluents From Accidental and Routine Reactor Releases For

the Purpose of Implementing Appendix I.

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide using an alternate approach.

General Compliance or Alternate Approach Assessment

Routine and accidental releases of radioactive liquid, heat, and chemical discharges to the Columbia River via the CGS cooling tower blowdown line are discussed in

Section 2.4.12. CGS final Environmental Report (ER) 6.1.1.1 describes in detail the advection/diffusion equations used in the near-field thermal analysis. This analysis provides dispersion characteristics, presented in ER 5.1, to 500 ft below the point of discharge. A simplified and conservative approach to estimating the far-field concentrations of routine releases is presented in ER 5.2.2. The affects of an accidental release of radioactive liquid to the ground within the CGS site area were investigated

and are discussed in Section 2.4.13.3.

Specific Evaluation Reference

See Sections 2.4.12 and 2.4.13.3 and Environmental Report Sections 5.1, 5.2.2, and 6.1.1.1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-187 Regulatory Guide 1.114, Revision 1, November 1976 Guidance to Operator at the Controls of a Nuclear Power Plant.

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

Plant administrative procedures implement the requirements of this regulatory guide.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-188 Regulatory Guide 1.115, Revision 0, March 1976 Protection Against Low-Trajectory Turbine Missiles

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of

construction permit applications docketed after November 15, 1976.

General Compliance or Alternate Approach Assessment

Extensive amounts of concrete used in the construction of CGS serve as radiation shielding and formidable barriers protecting essential systems from low trajectory

missiles.

Specific Evaluation Reference

See Section 3.5.1.3.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-189 Regulatory Guide 1.116, Revision 0, June 1976 Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical

Equipment and Systems

I Design and Construction Phase

Compliance or Alternate Approach Statement

CGS complied with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

The requirements for installation, inspection, and testing are specified in procurement

documents which require a quality assurance program in compliance with ANSI N45.2.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-190 Regulatory Guide 1.117, Revision 0, June 1976 Tornado Design Classification

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed after February 15, 1977.

General Compliance or Alternate Approach Assessment

Essential systems are protected from tornadoes by structures designed for design basis tornadoes (DBT). See Regulatory Guides 1.27 and 1.76.

Specific Evaluation Reference

See Sections 3.3.2.4 and 9.2.5.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-191 Regulatory Guide 1.118, Revision 0, June 1976 Periodic Testing of Electric Power and Protection Systems.

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since the construction permit for CGS

was issued prior to February 15, 1977.

General Compliance or Alternate Approach Assessment

Electric power and protection systems are tested periodically as specified in the Technical Specifications. As described in Section 13.5.2 , surveillance procedures have been prepared for periodic testing of these systems.

Specific Evaluation Reference

See the Technical Specifications.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-192 Regulatory Guide 1.120, Revision 0, June 1976 Fire Protection Guidelines for Nuclear Power Plants

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed after February 28, 1977. However, the NRC requested a reevaluation of the fire protection program of CGS and a comparison with the guidelines in Appendix A to Branch Technical Position APCSB 9.5-1, Guidelines for Fire Protection For Nuclear Power Plants, Docketed Prior to July 1, 1976. CGS complies with the intent of the guidance set forth in Appendix A to Branch Technical

Position APCSB 9.5-1.

General Compliance or Alternate Approach Assessment

Appendix F includes the fire hazard analysis and compares in detail the fire protection provisions for CGS with the guidelines in Appendix A to Branch Technical

Position APCSB 9.5-1.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-193 Regulatory Guide 1.122, Revision 0, September 1976 Development of Floor Design Response Spectra for Seismic Design of Floor Supported

Equipment or Components

Compliance or Alternate Approach Statement

CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

CGS complies with some of the regulatory positions and where not in compliance, alternate methods are used as discussed in Sections 3.7.2.5 and 3.7.2.6.

Specific Evaluation Reference

See Sections 3.7.2.5 and 3.7.2.6.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-194 Regulatory Guide 1.123, Revision 0, October 1976 Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear

Power Plants

I Design and Construction Phase

Compliance or Alternate Approach Statement

CGS complied with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

ANSI N45.2.13-1976, the subject of this regulatory guide, requires certain supplier selection, evaluation, and pre- and post-award activities.

Prequalification of suppliers was generally not performed. The procurement documents

required prospective suppliers to submit information pertaining to experience, facilities, personnel, and quality program with their bids for evaluation prior to award of a

contract.

Pre-award evaluations were restricted to the information submitted with bid and selected clarifications when an adequate evaluati on could not be accomplished with the information supplied. Post-award evaluations were performed in conjunction with the quality assurance program evaluation and approval after award of a contract.

Inspection and hold points were not established through agreement with the bidder but through contract requirements to notify Energy Northwest of all inspections and tests which were selectively witnessed by Energy Northwest.

Specific Evaluation Reference

None II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-195 Regulatory Guide 1.124, Revision 0, November 1976 Design Limits and Loading Combinations for Class 1 Linear Type Component Supports

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed after July 1, 1977. However, CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

Design and fabrication requirements for CGS, including those requirements for linear type components supports, are in accordance with the ASME Code Section III Subsection NF, Winter 1973 Addenda. The actual design criteria were established

prior to Winter 1973 Addenda and are conservative with respect to the Winter 1973 Code. Regulatory Guide 1.124 provides design limits and appropriate combinations of

loadings which reflect the requirements set forth in the 1974 Edition of the ASME Code Section III, Subsection NF, along with additional requirements. Although the detailed requirements of the regulatory guide have not been incorporated as project criteria, review of the design criteria used for CGS indicates that the intent of this regulatory

guide is met.

Specific Evaluation Reference

See Sections 3.9.3.4 and 5.4.14.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-196 Regulatory Guide 1.125, Revision 0, March 1977 Physical Models for Design and Operation of Hydraulic Structures and Systems for Nuclear

Power Plants

Compliance or Alternate Approach Statement

The guide is not applicable to CGS since it applies to the evaluation of construction permit application docketed on or after November 1, 1977. Furthermore, the guide is not applicable to CGS for reasons stated below.

General Compliance or Alternate Approach Assessment:

Physical hydraulic model testing is not used for CGS for predicting the performance of hydraulic structures, systems, and components located outside the primary containment vessel or provided for the prevention of accidents and the mitigation of the consequences of accidents. Therefore, the details and documentation of data and

studies required by the guide to support such testing is not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-197 Regulatory Guide 1.127, Revision 0, April 1977 Inspection of Water-Control Structures Associated With Nuclear Power Plants

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since water-control structures as defined

in this regulatory guide do not exist.

General Compliance or Alternate Approach Assessment

Not applicable.

Specific Evaluation Reference

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-198 Regulatory Guide 1.128, Revision 0, April 1977 Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants.

Compliance or Alternate Approach Statement

This regulatory guide is not applicable to CGS since it applies to the evaluation of construction permit applications docketed after December 1, 1977. However, CGS complies with the intent of the guidance set forth in this regulatory guide by an alternate approach.

General Compliance or Alternate Approach Assessment

Safety-related battery installation design criteria conforms to IEEE 484-1975.

A Class 1E ventilation system is also provided which is capable of limiting hydrogen concentrations to 1%.

Storage prior to installation was not in strict compliance with Section 5.1.3 of this regulatory guide. However, preoperational tests established whether or not any damage

or loss of capacity resulted from storage.

Specific Evaluation Reference

See Sections 8.3.2.1.5 , 8.3.2.1.6 , 8.3.2.2.1.1 , and 8.3.2.2.1.2.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-030 1.8-199 Regulatory Guide 1.129, Revision 0, April 1977

Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power

Plants

Compliance or Alternate Approach Statement

Although Regulatory Guide 1.129 is not directly applicable to CGS, Energy

Northwests maintenance procedures conform to IEEE 450-2002, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead

Storage Batteries for Generating Stations and Substations. The frequency for service testing is in accordance with Technical Specifications or Licensee Controlled Specifications.

General Compliance or Alternate Approach Assessment

See Section 8.3.2.1.7.

Specific Evaluation Reference

See Section 8.3.2.1.7.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-09-010 1.8-200 Regulatory Guide 1.137, Revision 1, October 1979

Fuel Oil Systems for Standby Diesel Generators

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this guide with the exception of the

following:

Piping on the engine skid is ANSI B 31.1, Seismic Category I, Quality Class I, as

noted in Section 9.5.4.1. Item 11, cathodic protection surveillance. The standby diesel fuel oil storage tanks are protected with cathodic protection by anodes which are located in the near vicinity, but there are no pigtails connected to the fuel oil system piping, thus no leads to maintain.

CGS does not perform the 90% distillation test before putting the fuel in the tanks as

noted in Section 9.5.4.4 and the Technical Specifications.

The diesel fuel oil supply is gravity feed down to the low fuel oil alarm level. The

pump suction, however, is 2.3 ft higher than the bottom of the tank. Therefore, if the

transfer pump fails, the last few hours of running before the day tank is empty would be at a pump suction lift of up to 2.3 ft.

The auxiliary boiler storage tank is considered part of the diesel fuel oil system in that

it is an additional diesel fuel oil storage tank. This deviates from the ANSI N195-1976 standard because of the permanent interconnection between the standby power system and the auxiliary boiler system. The auxiliary boiler storage tank and its connective

piping are not Safety Class 3. The auxiliary boiler storage tank and its connecting auxiliary boiler system are not in a vital area, although ANSI N195-1976 specifies that the fuel oil system is a vital system and shall be located in a vital area. However, loss

of the stored fuel oil in the auxiliary boiler storage tank or its connective piping will not affect the safety function of the diesel fuel oil system.

The diesel storage minimum required volume does not include volume for testing, as specified by ANSI N195-1976. Instead, Energy Northwest procedurally provides for makeup, as needed, during testing activities to ensure that the minimum required volume is maintained.

Specific Evaluation Reference

See Section 9.5.4.4.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-201 Regulatory Guide 1.143, Revision 1, October 1979 Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants

Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment

CGS began implementing the guidance set forth in this regulatory guide in July 1982. Prior to this time the solid, liquid, and gaseous radioactive waste systems were being designed and fabricated as ASME Section III, Class 3, systems. Therefore, although the guidance in the regulatory guide does not call for N-stamped components, in many

cases N-stamped components are found in the radwaste systems. To avoid the confusion which may result from the implementation of this regulatory guide these

systems, and components which follow the guidance found in the regulatory guide are

indicated as Quality Class II+ and Code Group D+.

Specific Evaluation

Reference:

See Sections 3.2.4 and 3.2.6.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-202 Regulatory Guide 1.144, Revision 1, September 1980 Auditing of Quality Assurance Programs for Nuclear Power Plants I Design and Construction Phase Compliance or Alternate Approach Statement

CGS complies with the guidance set forth in this regulatory guide as described below.

General Compliance or Alternate Approach Assessment:

Contractors and suppliers complied with the requirements imposed by procurement documents.

Energy Northwest, the architect-engineer (Burns and Roe), and the construction manager (Bechtel) complied with the guidance set forth in this regulatory guide except

for the following.

The requirements of ANSI N45.2.12-1977 as modified and interpreted by the regulatory position were applied to the Bechtel quality program for safety-related items except as modified or interpreted below:

a.

Reference:

Standard Sections 4.3.2.4 and 4.5.1 (Investigation). As an equivalent alternative to the requirement for the audited organization to investigate any adverse audit finding to determine and schedule appropriate corrective action, Bechtels auditing organization may determine the

investigatory action and corrective action including action to prevent recurrence pertinent to adverse audit finding. These actions are agreed to by the audited organization. Further, in Section 4.5.1, as equivalent alternative to the 30-day response time, a response time appropriate to the finding is agreed to by the

audited and auditing organizations.

b.

Reference:

Regulatory Section C.7, Standard Section 5.2 (Audit Records). Audit records shall include documents as defined in the standard and other

documents if necessary to support audit findings.

Early project procurements specified audit program requirements in terms of Appendix B to 10 CFR 50 and ANSI N45.2. As appropriate, future procurements required that audit programs comply with ANSI Standard N45.2.12.

II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-203 Regulatory Guide 1.145, Revision 1, November 1982/February 1983 Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear

Power Plants

Regulatory Guide Intent This guide provides acceptable methodology to determining site-specific relative concentrations for assessing the potential offsite radiological consequences of postulated accidental releases of radioactive material to the atmosphere.

Application Assessment Assessed capability in design.

Compliance or Alternate Approach Statement

Identified BOP scope of supply analysis, design, and/or equipment used in this facility

is in full compliance with the regulatory guide.

General Compliance or Alternate Approach Assessment Two of the procedures contained in the PAVAN code were implemented. The procedures were run with the desert sigma and with the Pasquill-Gifford sigma enabled. The most conservative /Q values were used in the accident analysis.

Specific Evaluation Reference

See Section 2.3 and Chapter 15.0.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-204 Regulatory Guide 1.146, Revision 0, August 1980 Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants

I Design and Construction Phase

Compliance or Alternate Approach Statement

CGS complied with the guidance set forth in this regulatory guide as described below.

General Compliance or Alternate Approach Assessment

Energy Northwest, the architect-engineer (Burns and Roe), and the construction manager (Bechtel) complied with the guidance set forth in this regulatory guide.

Contractors and suppliers comply with the requirements imposed by procurement

documents.

Early project procurements specified audit program requirements in terms of Appendix B 10 CFR 50 and ANSI N45.2. Where appropriate, future procurements required that auditor qualification comply with ANSI Standard N45.2.23.

II Operational Phase Compliance is discussed in the OQAPD.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-205 Regulatory Guide 1.147 Inservice Inspection of Code Case Acceptability ASME Section XI Division I.

By the reference below, the NRC approved application of Code Case N416 for CGS which at that time was not addressed in Regulatory Guide 1.147. The approval letter required that Energy Northwest document application of the code case in the FSAR.

The code case was first used for CGS in 1988 for deferral of hydrostatic testing of main steam drip line modifications.

As the code case has now been accepted by Regulatory Guide 1.147, Energy Northwest does

not plan to document future use of the code case.

Reference:

Letter from T. M. Novak (NRC) to G. C. Sorensen (SS), Use of ASME Code Case N-416 for the WNP-2, WPPSS Nuclear Project No. 2 (WNP-2), dated August 8, 1985.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 1.8-206 Regulatory Guide 1.155, Reissued August 1988 Station Blackout

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

Compliance or Alternate Approach Assessment:

Regulatory Guide 1.155 was issued to describe a method acceptable to the NRC staff for complying with the NRC regulation that requires nuclear power plants to be capable of coping with a station blackout for a specified duration. The NRC acceptance of the CGS proposed plan for providing this capability is provided in the reference.

Specific Evaluation

Reference:

See Appendix 8A.

Reference:

Letter from R. R. Assa to G. C. Sorensen, Supplemental Safety Evaluation (SSE) of

the Washington Public Power Supply System Nuclear Project No. 2 (WNP-2) Station

Blackout Analysis (TAC M68626), dated June 26, 1992.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-14-029 1.8-207 Regulatory Guide 1.160, Revision 3, May 2012

Monitoring the Effectiveness of Maintenance at Nuclear Power Plants

Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

Compliance with the guidance provided is ensured by the implementation of a

maintenance program and implementing procedures at CGS.

Specific Evaluation

References:

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-025 1.8-208 Regulatory Guide 1.196, May 2003 Control Room Habitability at Light-Water Nuclear Power Reactors Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

Compliance with the guidance provided is ensured by the implementation of a Control Room Envelope Habitability (CREH) Program and implementing procedures at CGS.

Specific Evaluation

References:

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-025 1.8-209 Regulatory Guide 1.197, May 2003 Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors Compliance or Alternate Approach Statement:

CGS complies with the guidance set forth in this regulatory guide.

General Compliance or Alternate Approach Assessment:

Compliance with the guidance provided is ensured by the implementation of a Control Room Envelope Habitability (CREH) Program and implementing procedures at CGS.

Specific Evaluation

References:

Not applicable.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 2 SITE CHARACTERISTICS

TABLE OF CONTENTS

Section Page LDCN-09-044 2-i 2.1 GEOGRAPHY AND DEMOGRAPHY ................................................ 2.1-1 2.1.1 SITE LOCATION AND DESCRIPTION ........................................... 2.

1-1 2.1.1.1 Specificati on of Location ............................................................. 2.

1-1 2.1.1.2 Site Area Map

.......................................................................... 2.1-1 2.1.1.3 Boundaries for Establishing Effluent Release Limits ........................... 2.1-2 2.1.2 EXCLUSION AREA AUTHORITY AND CONTROL .......................... 2.1-2 2.1.2.1 Aut hority ................................................................................

2.1-2 2.1.2.2 Control of Activities Unrelated to Plant Op eration .............................. 2.1-4 2.1.2.2.1 Industrial Development Complex ................................................ 2.1-4 2.1.2.2.2 618-11 (Wye) Waste Burial Gr ound .............................................

2.1-4 2.1.2.3 Arrangements for Traffic Cont rol ..................................................

2.1-5 2.1.2.4 Abandonment or Relocation of Roads ............................................. 2.1-5

2.1.3 POPULATION

DISTRIBUTION ..................................................... 2.

1-5 2.1.3.1 Population W ithin Ten M iles ........................................................

2.1-6 2.1.3.2 Population Betw een Ten and Fifty Miles

.......................................... 2.1-7 2.1.3.3 Transient Populati on ..................................................................

2.1-7 2.1.3.4 Low Population Zone ................................................................. 2.1-7 2.1.3.5 Populati on Center

..................................................................... 2.1-7 2.1.3.6 Populati on Density

.................................................................... 2.1-7 2.

1.4 REFERENCES

........................................................................... 2.1-8

2.2 NEARBY

INDUSTRIAL, TR ANSPORTATION, AND MILITARY FACILITIES ................................................................................ 2.2-1

2.2.1 LOCATION

AND ROUTES .......................................................... 2.2-1 2.

2.2 DESCRIPTION

S

......................................................................... 2.2-3 2.2.2.1 Description of Facilities .............................................................. 2.2-3 2.2.2.2 Description of Products and Materials

............................................. 2.2-6 2.2.2.3 Pipe lines ................................................................................

2.2-7 2.2.2.4 Wate rways ..............................................................................

2.2-7 2.2.2.5 Airports ................................................................................. 2.2-8 2.2.2.6 Projection of Industrial Growth ..................................................... 2.2-8

2.2.3 EVALUATION

OF PO TENTIAL ACCIDENTS ................................. 2.2-8 2.2.3.1 Determination of Design Basis Events

............................................. 2.2-8 2.2.3.2 Effects of Design Basis Events

...................................................... 2.2-11 2.

2.4 REFERENCES

........................................................................... 2.2-11

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 2 SITE CHARACTERISTICS

TABLE OF CONTENTS (Continued)

Section Page 2-ii 2.3 METEOROLOGY ......................................................................... 2.3-1

2.3.1 REGIONAL

CL IMATOLOGY

.......................................................... 2.3-1 2.3.1.1 Genera l Climate

....................................................................... 2.3-1 2.3.1.2 Regional Mete orological Conditions for Design and Operating Bases

....... 2.3-4 2.3.1.2.1 Severe Weat her Phenomena ....................................................... 2.

3-4 2.3.1.2.1.1 Heavy Ra in, Snow, and Ice .....................................................

2.3-4 2.3.1.2.1.2 Thunderstorms and Hail

......................................................... 2.3-4 2.3.1.2.1.3 To rnadoes .......................................................................... 2.3-6 2.3.1.2.1.4 Strong Winds

...................................................................... 2.3-7 2.3.1.2.1.5 High Air Pollution Potential (APP) and Dust Storm Potential ........... 2.3-8 2.3.1.2.1.5.1 Evaluation of August 11, 1955 and January 11, 1972 Dust Storms at Hanford ............................................................. 2.3-11 2.3.1.2.1.5.2 Hanf ord Dust Storm Climatology for Design and Operating Bases ... 2.3-13 2.3.1.2.2 Design Snow Load ..................................................................

2.3-14 2.3.1.2.3 Meteorological Data Used for Evaluation of Ultimate Heat Sink .......... 2.3-15

2.3.2 LOCAL

METEOROLOGY .............................................................. 2.3-16 2.3.2.1 Data Comparisons

..................................................................... 2.3-16 2.3.2.1.1 Winds

.................................................................................. 2.3-18 2.3.2.1.2 Moisture and Temperature ........................................................ 2.3-19 2.3.2.1.3 Monthly Precipitation

.............................................................. 2.3-20 2.3.2.1.4 Fog .................................................................................... 2.3-20 2.3.2.1.5 Stability Summaries

................................................................. 2.3-21 2.3.2.2 Potential Influence of the Plant and Its Facilities on Local Meteorology .... 2.3-22 2.3.2.3 Local Meteorological Conditions for Design and Operating Bases

........... 2.3-23 2.3.2.4 Topographic Description

............................................................. 2.3-23

2.3.3 ONSITE

METEOROLOGICAL MEASUREMENT PROGRAM ................... 2.3-23 2.3.3.1 Permanent Onsite Meteorological Tower and Instrumentation Characteristics ......................................................................... 2.

3-24 2.3.3.2 Quality Assurance Program ......................................................... 2.3-26 2.3.3.2.1 Data Recovery During Apr il 1, 1974 - March 31, 1976 ..................... 2.3-26 2.3.3.2.2 Maintenance and Calibration

..................................................... 2.3-28 2.3.3.2.3 Data Proce ssing and Analysis .................................................... 2.3-28 2.3.3.2.4 Meteorological Monitoring Program During Plant Operation .............. 2.3-29 2.3.3.3 Other Meteorological Measurement Programs Considered for the Data Comparisons ............................................................................ 2.3-31 2.3.3.3.1 CGS Tem porary Tower

............................................................ 2.3-31 C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 2 SITE CHARACTERISTICS

TABLE OF CONTENTS (Continued)

Section Page 2-iii 2.3.3.3.2 Hanford Mete orological Sta tion ..................................................

2.3-31 2.3.3.4 Joint Stability - Wind Frequency Summaries ..................................... 2.3-31

2.3.4 SHORT

TERM DIFFUSI ON ESTIMATES ........................................ 2.3-32 2.3.4.1 Objective ................................................................................

2.3-32 2.3.4.2 Exclusion Area Boundary

............................................................ 2.3-32 2.3.4.3 Low Population Zone

................................................................. 2.3-33 2.3.4.4 Control Room .......................................................................... 2.3-33 2.3.4.5 Description of Sources

............................................................... 2.3-33 2.3.4.6 Control Ro om Intakes

................................................................ 2.3-34 2.3.4.7 Calculations ............................................................................ 2.3-34 2.3.5 LONG-TERM (ROUTINE)

DIFFUSION ESTIMATES ............................. 2.3-35 2.3.5.1 Objectives ............................................................................... 2.3-35 2.3.5.2 Calcul ations ............................................................................

2.3-36 2.

3.6 REFERENCES

........................................................................... 2.3-38

2.4 HYDROLOGY

EN GINEERING

........................................................ 2.4-1

2.4.1 HYDROLOGIC

DESCRIPTION ..................................................... 2.4-1 2.4.1.1 Site and Facilities ...................................................................... 2.4-1 2.4.1.2 Hydrosphere ............................................................................ 2.4-1

2.4.2 FLOODS

.................................................................................. 2.4-4 2.4.2.1 Flood History .......................................................................... 2.4-4 2.4.2.2 Flood Design C onsiderations ........................................................ 2.4-4 2.4.2.3 Effects of Local Intense Precipitation ..............................................

2.4-5 2.4.3 PROBABLE MAXIMUM FLOOD ON STREAMS AND RIVERS ........... 2.4-6 2.4.3.1 Probable Maxi mum Precipitation ................................................... 2.4-7 2.4.3.2 Precipita tion Losses

................................................................... 2.4-7 2.4.3.3 Runoff and Stream Course Models ................................................. 2.4-7 2.4.3.4 Probable Maxi mum Flood Fl ow ....................................................

2.4-8 2.4.3.5 Water Le vel Determinations

......................................................... 2.4-9 2.4.3.6 Coincident Wi nd Wave Activity .................................................... 2.4-9

2.4.4 POTENTIAL

DAM FAILURES, SEISMICALLY INDUCED ................. 2.4-9 2.4.4.1 Dam Failure Permutations ........................................................... 2.4-10 2.4.4.2 Unsteady Flow Analysis of Potentia l Dam Failures ............................. 2.4-11 2.4.4.3 Water Le vel at Plant Site

............................................................. 2.4-11 2.4.5 PROBABLE MAXIMUM SURGE AND SEICHE FLOODING ............... 2.4-11 2.4.6 PROBABLE MAXIMUM TSUNAMI FLOODING .............................. 2.4-12

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 Chapter 2 SITE CHARACTERISTICS

TABLE OF CONTENTS (Continued)

Section Page 2-iv 2.4.7 ICE E FFECTS ...........................................................................

2.4-12 2.4.8 COOLING WATER CANALS AND RESERVOIRS ............................. 2.4-13

2.4.9 CHANNEL

DI VERSIONS ............................................................ 2.4-13 2.4.10 FLOODING PROTECTI ON REQUIREMENTS ................................ 2.4-13 2.4.11 LOW WATER CONS IDERATIONS

.................................................. 2.4-14 2.4.11.1 Low Flow in Streams

................................................................ 2.4-14 2.4.11.2 Low Water Resulting From Sur ges, Seiches, or Tsunami ..................... 2.4-15 2.4.11.3 Historical Low Water

............................................................... 2.4-15 2.4.11.4 Future Controls

....................................................................... 2.4-16 2.4.11.5 Plant Re quirements

.................................................................. 2.4-17 2.4.11.6 Heat Sink Dependa bility Requireme nts ..........................................

2.4-17 2.4.12 DISPERSION, DILUTION, AND TRAVEL TIMES OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS ........... 2.4-17 2.4.13 GROUNDWATER ..................................................................... 2.4-18 2.4.13.1 Description and Onsite Use ........................................................

2.4-18 2.4.13.2 S ources ................................................................................

2.4-23 2.4.13.3 Accident al Effects

.................................................................... 2.4-25 2.4.13.4 Monitoring or Sa feguard Require ments ..........................................

2.4-29 2.4.13.5 Design Bases for Subsur face Hydrostatic Lo adings ........................... 2.4-29 2.4.14 TECHNICAL SPECIFICATIO NS AND EMERGENCY OPERATION REQUIREMENTS ..................................................................... 2.4-30 2.4.15 REFERENCES ......................................................................... 2.4-30 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING ...... 2.5-1

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF TABLES

Number Title Page 2-v 2.1-1 Projected Population Distribution by Compass Sector and Distance from the Site.....................................................................2.1-9 2.1-2 2000 Populati on Distribution by Compass Sector and Distance from the Site..........................................................2.1-12

2.2-1 Hanford Site Nuclear Facilities...............................................2.2-13

2.3-1 Average and Extremes of Climatic Elements at Hanford................2.3-43

2.3-2 Average Return Period a nd Existing Record for Various Precipitation Amounts and Intensity During Specified Time Periods at Hanford..............................................................

2.3-45 2.3-3 Miscellaneous Snowfall Statistics (1946 through 1970)..................2.3-46

2.3-4 Tornado History Within 100 Miles of CGS................................2.3-47

2.3-5 Monthly and Annual Prevailing Directions, Average Speeds, and Peak Gusts: 1945-1970 at HMS........................................2.3-48

2.3-6 Speed and Direction of Daily Peak Gusts..................................

2.3-49 2.3-7a CGS and HMS Hourly Mete orlogical Data, August 7-9, 1972 (Ultimate Heat Sink Studies)..................................................2.3-51

2.3-7b CGS Hourly Meteorolog ical Data, July 4-12, 1975 (Ultimate Heat Si nk Studies)..................................................2.3-52

2.3-7c CGS Hourly Meteorolog ical Data, July 4-12, 1975 (Ultimate Heat Si nk Studies)..................................................2.3-53

2.3-7d CGS Hourly Meteorolog ical Data, July 4-12, 1975 (Ultimate Heat Si nk Studies)..................................................2.3-54

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF TABLES (Continued)

Number Title Page 2-vi 2.3-7e CGS Hourly Meteorolog ical Data, July 4-12, 1975 (Ultimate Heat Sink Studies)..................................................2.3-55 2.3-7f 24 Hour HMS Mete orological Profile for August 4, 1961..............2.3-56

2.3-7g Diurnal Varia tion in Dry Bulb and Wet Bulb Temperature for Use in Analyzing Second Through Thirtieth Day Pond Thermal Performance..........................................................

2.3-57 2.3-7h Diurnal Varia tion in Dry Bulb and Wet Bulb Temperature for Use in Analyzing First Th rough Thirtieth Day Maximum Mass Loss........................................................................2.

3-58 2.3-8a Summary of CGS Onsite Meteorological Data Collected During the First and Second Annual Cycles as Compared to Corresponding Hanford Meterological Sta tion Data......................2.3-59

2.3-8b Frequency of Occurrence of Wind Direction Versus Speed for CGS 33-ft Level (1974-1975).................................................2.3-61

2.3-9 Percentage Fre quency Distribution of 50-ft Wind Direction Versus Speed at HM S (1955-1970)..........................................2.3-66

2.3-10 Percent Frequency of Occu rrence of Wind Direction at the Hanford Reserv ation............................................................

2.3-70 2.3-11 Persistence of Wind Direction in One Se ctor (22.5 Degrees) from 4/74 through 3/75 at 33-ft Level......................................2.3-72

2.3-12 Persistence of Wind Direction in Two Sectors (45 Degrees) from 4/74 through 3/75 at CGS for 33-ft Level...........................2.3-74

2.3-12a Longest Persiste nce of Wind Direction in One (22.5 Degrees) and Two (45 Degrees) Sector s During First and Second Annual Cycles at 33-ft Level.................................................2.3-76

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF TABLES (Continued)

Number Title Page 2-vii 2.3-13 Percent Frequency of Occurrence of Wind Speed at the Hanford Reserv ation............................................................

2.3-77 2.3-14 Diurnal Variation of 33-ft El evation Dry Bulb Temperature (F°)

at CGS and Monthly Average Dry Bulb Temperature (F°) at the Hanford Reservation.......................................................2.3-79

2.3-15 Diurnal Variation of 33-ft Elevation Wet Bulb Temperature (F°)

at CGS and Monthly Average Wet Bulb Temperature (F°) at the Hanford Reserv ation............................................................

2.3-80 2.3-16 Diurnal Variation of 33 ft Elevation Dew Point Temperature (F°)

at CGS and Monthly Average Dew Point Temperature (F°) at the Hanford Reservation.......................................................2.3-81

2.3-17 Frequency of Occurrence, Dry Bulb Temperature (F°) Versus Time of Day from 4/74 through 3/75 for 33-ft Level....................2.3-82

2.3-18 Frequency of Occurrence, Wet Bulb Temperature (F°) Versus Time of Day from 4/74 through 3/75 for 33-ft Level....................2.3-83

2.3-19 Frequency of Occurrence, Dew Point Temperature (F°) Versus Time of Day from 4/74 through 3/75 for 33-ft Level....................2.3-84

2.3-20 Monthly Averages of Psyc hrometric Data Based on Period of Record (1950-1970).........................................................

2.3-85 2.3-21 Diurnal Variation of Pr ecipitation Intensity at CGS and Monthly Total Precipitation at th e Hanford Reservation.................2.3-86

2.3-22 Frequency of Occurrence, Pr ecipitation Versus Time of Day from 4/74 through 3/

75 at CGS..............................................2.3-87

2.3-22a Annual Frequency of Occu rrence of Wind Direction and Wind Speed Versus Precip itation Intensity.................................2.3-88

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Chapter 2 SITE CHARACTERISTICS

LIST OF TABLES (Continued)

Number Title Page LDCN-13-067 2-viii 2.3-23 Statistics on Fog at the Hanford Meteorology Station .................... 2.3-90 2.3-24 Percent Freque ncy Distribution of Wi nd Speeds During Hourly Observations of Fog at Pasco and at HMS ................................. 2.3-91 2.3-25 Percent Frequency of O ccurrence of Stability at the Hanford Reserva tion............................................................ 2.3-92 2.3-26 Frequency of Occurrence T ( F/200 ft) Versus Time of Day from 4/74 through 3/75 at CGS between 245 and 33-ft Levels ......... 2.3-94

2.3-27 Frequency of Occurrence, Sigma () Versus Time of Day from 4/74 through 3/75 at CGS fo r 33-ft Level .................................. 2.3-95 2.3-28 Joint Frequency Distribution of Wind Speed and Direction ............. 2.3-96

2.3-28A Joint Frequency Di stribution of Wind Speed and Direction ............. 2.3-100

2.3-29 through 2.3-32 DELETED

2.3-33 Exclusion Ar ea Boundary Accident /Q Desert Sigmas ................. 2.3-105

2.3-33a Exclusion Area Boundary /Q Values Desert Sigmas w/ Meander .... 2.3-106

2.3-34 Exclusion Ar ea Boundary Accident /Q P-G Sigmas .................... 2.3-107 2.3-34a Exclusion Area Boundary /Q Values Pasquill-Gifford Sigmar w/ Meander and Building Wake Credit ..................................... 2.3-108 2.3-35 Low Population Zone Accident /Q Desert Sigmas ...................... 2.

3-109 2.3-36 Low Population Zone Accident /Q P-G Sigmas ......................... 2.3-110 2.3-37 Control Room, Excl usion Area Boundary and Low Population Zone /Qs (S/m 3) .......................................... 2.

3-111 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Chapter 2 SITE CHARACTERISTICS

LIST OF TABLES (Continued)

Number Title Page LDCN-13-067 2-ix 2.3-38a CGS Calculation, Terrain Features, Desert Sigmas ...................... 2.3-112

2.3-38b CGS Calculation, Terrain F eatures, Desert Sigmas ...................... 2.3-113

2.3-38c CGS Calculations, Terrain Features, Desert Sigmas ..................... 2.3-114

2.3-38d CGS Calculation, Terrain F eatures, Desert Sigmas ...................... 2.3-115

2.3-38e CGS Calculations, Terrain Features, Desert Sigmas ..................... 2.3-116

2.3-38f CGS Calculation, Terrain Features, Desert Sigmas ...................... 2.3-117

2.3-39a DELETED ....................................................................... 2.3-118 through 2.3-39f

2.3-40 Frequency of Wind Resuspension Pe riods at Hanford (1953-1970)

...................................................................... 2.3-119 2.3-41 Dust Concentration Depende ncy on Wind Speed and Direction at Hanford 1953-1970 .......................................................... 2.

3-120 2.3-42 Hours Satisfying Dust Storm Cr iteria at Hanford (1953-1970) ......... 2.3-121

2.4-1 Major Columbia River Basin Dams ......................................... 2.4-35

2.4-2 Columbia River Temperatures Near Columbia Generating Station .... 2.4-36

2.4-3 Downstream Surface Water Users

........................................... 2.4-37

2.4-4 Mean Discharges in CFS of Columbia River Below Priest Rapids Dam, Modified to 1970 Conditions ................................ 2.4-38

2.4-5 Dependable Yield, Columbia River Below Priest Rapids Dam, Washington

....................................................................... 2.4-39 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF TABLES (Continued)

Number Title Page LDCN-05-050 2-x 2.4-6 Major Geologic Units in the Hanford Region and Their Water-Bearing Pr operties......................................................2.4-40

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF FIGURES

Number Title 2-xi 2.1-1 General Area, 0-20 Miles 2.1-2 General Area, 0-50 Miles

2.1-3 Overall Site Plan

2.1-4 Project Area Map - 10 Mile Radius

2.1-5 Project Area Map - 50 Mile Radius

2.1-6 2000 Population in Communities Around Site

2.1-7 Transportation and Topographic Features of Low Population Zone

2.2-1 Hanford Reservation Road System

2.2-2 Hanford Reserva tion Railroad System

2.2-3 Federal Airways and Inst rument Approaches/Departures

2.3-1 Rainfall Intensity, Duration, a nd Frequency Based on the Period 1947-69 at Hanford

2.3-2 Greatest Depth of Snow on Ground During 24 of 25 Winters of Record at Hanford 1946-47 through 1969-70

2.3-3 Distribution of Characterized To rnadoes in 20-Year Period (1950-69) 2.3-4 Peak Wind Gust Return Probability Diagram

2.3-5 Dust Occurrences Per Wind Speeds to 400 ft Heights

2.3-6 Near-Surface Airborne Dust Concentration as a Function of Average Air Velocity C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF FIGURES (Continued)

Number Title 2-xii 2.3-7 Monthly Hourly Average of Temperature and Rela tive Humidity (Sheets 1 through 3) 2.3-8 Probability (%) that the First Hourly Observation of an Inversion will Mark the Beginning of an Inversion Run of N Hr (Sheets 1 through 4)

2.3-9 Topographic Cross Sections of Region Surrounding Site

2.3-10 Yearly Hourly Average of Temperature and Relative Humidity 2.3-11 Cumulative (%) Frequency of Hourly Centerline /Q at Site Boundary Circular Distance of 1.212 Miles From Source (April 1974-March 1975)

2.3-12 Cumulative (%) Frequency of Hourly Centerline /Q at Site Boundary Circular Distance of 1.212 Miles From Source (April 1975-March 1976)

2.3-13 Cumulative (%) Fre quency of Occurrence of /Q For Postulated Accidents of 8, 16, 72, and 624 Hr At Ou ter Boundary of LPZ (3

.0 Miles from Source) (April 1974-March 1975)

2.3-14 Cumulative (%) Fre quency of Occurrence of /Q for Postulated Accidents of 8, 16, 72, and 624 Hr at Outer Boundary of LPZ (3.0 Miles from Source) (April 1975-March 1976)

2.3-15 Annual Average /Q by Sector at the Site Boundary for First and Second Annual Cycle Data 2.4-1 Hydrographic Map

2.4-2 Topographic Map of Site and Surrounding Area

2.4-3 Detailed Contours Near the Site

2.4-4 Discharge and Temperature of th e Columbia River at Priest Rapids

2.4-5 Columbia River Water Surface Profiles River Miles 323 to 358 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF FIGURES (Continued)

Number Title LDCN-03-058 2-xiii 2.4-6 Safety-Related Building Roof Plan s and Sections (Sheets 1 through 5) 2.4-7 PMF Hydrograph due to Thunderstorm PMP

2.4-8 Probable Maximum Preci pitation Drainage Basin

2.4-9 Probable Maximum Precipitation Channel Cross Sections

2.4-10 Water Surface Profile

2.4-11 Effective Fetch Diagram

2.4-12 Computed Long-Term Temperature Trends on the Columbia River at Rock Island Dam (1938-1962)

2.4-13 River Elevation at Low Flows - River Mile 352

2.4-14 Duration Curves Columbia River, Priest Rapids Dam

2.4-15 Frequency Curves of High and Low Flows for the Columbia River Below Priest Rapids Dam

2.4-16 Location of In take and Discharge

2.4-17 Monitoring Well Lo cations (September 1975)

2.4-18 Hanford Rese rvation Water Table Map (December 1975) 2.4-19 Groundwater Contour s Assuming Construction of the Ben Franklin Dam 2.4-20 Hanford Rese rvation Water Table Map (January 1944)

2.4-21 Hydraulic Conductivities in the Unconfined Aquifer

2.4-22 Nitrate (NO

3) Concentrations (December 1976)

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 2 SITE CHARACTERISTICS

LIST OF FIGURES (Continued)

Number Title LDCN-05-050 2-xiv 2.4-23 Gross Beta Concentrations (December 1976) 2.4-24 Tritium (3H) Concentrations (December 1976) 2.4-25 Pasco Basin Uppermost Confined Aquifer Potential Map (1970)

2.4-26 Hanford Reserv ation Water Table Map (September 1973)

2.4-27 Well Hydrographs (Sheets 1 and 2)

2.4-28 Site Topographic Map

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDCN-03-023 2.1-1 Chapter 2 SITE CHARACTERISTICS

2.1 GEOGRAPHY

AND DEMOGRAPHY

2.1.1 SITE LOCATION AND DESCRIPTION

2.1.1.1 Specification of Location

Columbia Generating Station (CGS) is located in the southeast area of the U.S.

Department of Energy's (DOE) Hanford Site in Benton County , Washington. The site is approximately 3 miles west of the Columbia River at River Mile 352, approximately 10 miles north of north Richland, 18 miles northwest of Pasco, and 21 miles northwest of Kennewick (Figures 2.1-1 and 2.1-2).

The reactor is located at 46° 28' 18" north latitude and 119° 19' 58" west longitude. The approximate Universal Transver se Mercator coordinates ar e 5,148,840 meters north and 320,930 meters east.

2.1.1.2 Site Area Map

The CGS site area is that real estate over which Ener gy Northwest has the legal right to control access. It is the ar ea enclosed by the exclusion area boundary plus the plant property lines as shown in Figure 3-1 of the Offsite Dose Calc ulation Manual (ODCM). The property line and nearby industrial facilities are shown in Figure 2.1-3. Industrial facilities located in the site area are the H. J. Ashe Subs tation and Energy Northwest's Nuclear Projects 1 and 4 (WNP-4 was terminated in January 1982, and WNP-1 was terminated in May 1994). Highway and railway facilities located within the site area are shown in Figure 2.1-3. The relative locations of the plant structures are shown in Figure 1.2-1.

The boundary of the exclusion area is a circle with its center at the reactor and a radius of 1950 m. Ownership and control of the land outside the CGS property line but within the site exclusion area are discussed in Section 2.1.2. The site is situated near the middle of the relatively flat, essentially featureless plain, which is best described as a shrub ste ppe with sagebrush interspersed with perennial native and introduced annual grasses extending in a northerly, westerly, and southerly direction for several miles. The plain is ch aracterized by slight topographic relief of approxi mately 20 ft across the plant site.

The dominant topographic features in the area are the Rattlesnake Hills, 13 to 15 miles west southwest, 3200 ft above the elevation of the plant site; Gable Mountain, approximately

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-13-050 2.1-2 10 miles northwest of the site a nd about 670 ft above the site grade; and the steep river cut bluffs forming the east bank of th e Columbia River, approximately 3.5 miles east of the site (Figure 2.1-1

). 2.1.1.3 Boundaries fo r Establishing Effluent Release Limits The boundary for establishing efflue nt release limits (unrestricted area boundary as defined in 10 CFR Part 20) is the site area boundary as shown in the ODCM, Figure 3-1. The site area is the area enclosed by the exclusion area boundary and the plant property lines that fall outside the exclusion area. All area with in the site area boundary is considered a controlled area as defined by 10 CFR 20.1003.

A number of restricted areas (as defined in 10 CFR 20.1003) are associated with CGS. The primary CGS restricted area is located within the plant security fence which also is the boundary of the protected area (a s defined in 10 CFR 73.2). This is shown as the double fence line in Figure 1.2-1. Unescorted access to the protected area is controlled by CGS security staff. Other restricted areas include the Independent Spent Fuel Storage Installation, stormwater ponds, Plant Support Facility calibration laboratory, Warehouse No. 5, the cooling tower sediment disposal area, and Building 167 on the WNP-4 site. Access to these secondary restricted areas is controlle d by locks and fences. Temporary restricted areas may be established and removed as dictated by activities at CGS.

2.1.2 EXCLUSION

AREA AUTHORITY AND CONTROL 2.1.2.1 Authority Energy Northwest leased 1089 acres from the DOE, within the DO E Hanford Site, to be used for CGS. A letter from the DOE Richland Oper ations office to the Managing Director of Energy Northwest (Reference 2.1-1) advises that the DOE has the authority to sell or lease land on the Hanford Site and the letter further states This Authority is contained in Section 120 of the Atomic Energy Community Act of 1955, as amended, and Section 161g of the At omic Energy Act of 1954, as amended. There is also general federal disposal authority available under the Federal Property Administrative Se rvices Act of 1949, as amended.

The 1950-m radius exclusion area extends beyo nd the CGS property lines and overlaps DOE lands as well as the additional land leased by Energy Northwest for the construction of the WNP-1 and WNP-4 projects (see Figure 2.1-3 and ODCM Figure 3-1). All land outside the Energy Northwest leased propert y but within the exclusion area is managed by the DOE.

In recognition of the requirement specified in 10 CFR 100.3(a) [Now 100.3] that a licensee have control over access to the exclusion area, the following terms have been incorporated as Article 7 of the site property lease agreement between Energy Northwest and the DOE (as modified in 1975):

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-11-027 2.1-3 Nothwithstanding any provisions of this lease to th e contrary, the Administration [Energy Research and Development Administration -- now DOE] agrees that the Supply System [now Energy Northwest]

has the authority to determine all activitie s within the exclusion area within the meaning of 10 CFR Section 100.3(a) [Now 100.3], including the authority to remove a personnel and property from the area. The Supply System agrees that it will exercise such authority in a manner so as not to preclude the Administration from undertaking any action or activity within the exclusion area that is permissible under the provisions of 10 CFR Section 100.3(a) [Now 100.3]. As used herein, the term "exclusion area" includes both the leas ed and nonleased portions of the exclusion area.

Therefore, any actions such as public access and actions concerning mineral rights and easements taken within the exclusion area but out side the leased property are under the control of the DOE with the provision that Energy Northwest has the legal right to control access of individuals to the exclusion area if necessary.

All rail shipments on the track which traverses the property (Figure 2.1-3) are also under control of the DOE and are also subject to the above provision and controls imposed by Energy Northwest Security.

The only paved roads that traverse the exclusion area of CGS are the CGS, WNP-1, and WNP-4 facility access roads shown in Figure 2.1-3. Access by land from outside of the Hanford Site to the plant site is over DOE roads. Travel within the exclusion area on the access roads will be under the authority of Energy Northwest.

In the event that evacuation or other control of the exclusi on area should become necessary, appropriate notice will be given to the DOE-Richland Operations Office for control of non-Energy Northwest orig inated activities.

The above provisions provide the necessary assu rance that the exclus ion area is properly controlled. If Energy Northwest should decide th at an easement would be useful in ensuring continued control, there is a provision in Article 5(b) of th e lease as follows:

Subject to the provisions of Section 161 g of the Atomic Energy Act of 1954, as amended, the Commission has authority to grant easements for rights-of-way for roads, transmission lines and for any other purpose, and agrees to negotiate with Energy Northwest for such rights-of-way over the Hanford Operations Area as are necessary to service the Leased Premises.

Pursuant to this provision Energy Northwest could obtain an easement over the exclusion area in question from the DOE, which would ensure th at no permanent structures or other activities inconsistent with the exclusion area would be carried on therein.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-038 2.1-4 2.1.2.2 Control of Activities Unrelated to Plant Operation

In accordance with, and as defined by 10 CFR 100.3, Energy Northwest has the authority to determine all activities within the exclusion area, including the authority to remove all personnel and property from the area. The following activities unrelated to plant operation are

permitted within the exclusion area:

2.1.2.2.1 Industrial Development Complex Energy-Northwest is conducting site restoration and economic deve lopment (such as leasing of excess facilities for office spac e and manufacturing) activities at the WNP-1 and WNP-4 sites (the WNP-1 and WNP-4 sites are also leased from the DOE and controlled by Energy Northwest). The number of personnel at the WNP-1 and WNP-4 sites varies. However, coordination of activities within the exclusion area is under th e control of Energy Northwest and the CGS emergency plan. This includes notif ication and evacuation considerations in the event of an emergency at CGS.

2.1.2.2.2 618-11 (Wye) Waste Burial Ground

The 618-11 site is a DOE waste burial ground, encompassing an eight-acre parcel directly adjacent to Energy Northwest leased land (see Figure 2.1-3) and located who lly within the CGS exclusion area. All 618-11 site activities are controlled by DOE in accordance with 10 CFR Chapter III. DOE has responsibility for the 618-11 site in accordance with 10 CFR 830.204. The soil overburden covering the caissons and vertical pipe units at the 618-11 site is identified as a passive design feature that se rves a mitigative function. Existing soil overburden shall not be removed.

Following non-intrusive surveillanc e and characterization activities, the site was returned to inactive status.

2.1.2.3 Arrangements for Traffic Control

The only roads within the exclus ion area are the Energy Northwest access roads. These roads are normally used only by employees and visitors associated with the CGS, WNP-1, and WNP-4 facilities, DOE, and DOE cont ractors. The secur ity force, with o ffsite assistance as required, controls tra ffic during emergencies.

2.1.2.4 Abandonment or Relocation of Roads

There were no public roads transversing the exclusion area that had to be abandoned or relocated as a result of the construction of CGS.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 2.1-5 2.1.3 POPULATION DISTRIBUTION

Table 2.1-1 presents the compass sector population estimates for 198 0 and the forecasts for the same compass sectors by decade from 1990 to 2030.

• Cumulative totals are also shown in Table 2.1-1. This table may be keyed to Figures 2.1-4 and 2.1-5 , which show the sectors and major population centers within 10 and 50 miles of the site. As can be seen in Figure 2.1-6 , population centers, within 50 miles of the site in clude the Tri-Cities area of Richland, Pasco, and Kennewick; Moses Lake; Herm iston; and the communities lying along the Yakima River from Prosser to Toppenish.

Figure 2.1-4 shows that there are no towns located within 10 miles of the site, with the excep tion of a small part of Richland.

The 1990 to 2030 forecasts pr esented here (Reference 2.1-2) are based on

a. 1979 population figures provided by the Washington State Office of Financial Management,

b. Benton and Franklin C ounty Traffic Analysis Zone population distributions,

c. Computed annual average area growth rates from 1975 through 1979 which were utilized to obtain the total 1980 population estimated for each area, and

d. County forecasts prepared by the Bonneville Power Administration. (References 2.1-3 and 2.1-4).

Table 2.1-2 presents the compass sect or population estimates for 2010 based on U.S. Census Bureau data (Reference 2.1-5). See also Figures 2.1-4 and 2.1-5.

2.1.3.1 Population Within Ten Miles

An estimated 3000 people live within 10 miles of the site. The near est inhabitants occupy farms which are located east of Columbia River and are thinly spread over five compass sectors. There are no permanent inhabita nts located within 3 miles of the site.

No significant changes in land use within five miles are anticipated.

The Hanford Site is expected to remain dedicated primarily to industrial use without private residences. No change in the use of the land east of the Columbia Rive r is expected since it currently is irrigated to

• Population estimates out to 50 miles were derived to serve the licensing requirements of WNP-1, CGS, and WNP-4. Therefore, estimates were made relative to the centroid of the triangle formed by the three reac tors. This point is locate d 2800 ft east of CGS and has coordinates longitude 11 9º 19' 18" west, latitude 46° 28' 19" north. This shift does not affect the overall accuracy or applicability of the population distribution projections.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 2.1-6 about the maximum amount practicable. Th e primary increase in population within the 10-mile radius is expected to be in the area south and south-southeast of the plant (see Figure 2.1-4

). 2.1.3.2 Population Betwee n Ten and Fifty Miles

As indicated in Table 2.1-2 , about 450,000 people were estimated to be living within a 50-mile radius of CGS in 2010. Projecti ons for the 10-50 mile region are shown in Table 2.1-1 which is based on earlier (1979-1980) population counts.

2.1.3.3 Transient Population

The transient population consists of agricultural worker s needed for harvesting crops produced in the region, industrial and construction workers, and sportsmen engaged in hunting, fishing, and boating. A description of the transient population is di scussed in Section 5.6 of the CGS Emergency Plan.

2.1.3.4 Low Population Zone

The low population zone (LPZ) [see 10 CFR 100.

3(b)] for CGS is defined as all land within a

3-mile radius of the reactor. This LPZ was selected on the basis th at it is not expected to have a large population in the future and that effective protective measures could be established. As shown in Table 2.1-2, no permanent residents are located within a 3-mile radius of the reactor, and none are anticipated in the future.

There are no public faciliti es or institutions such as schools a nd hospitals within a 3-mile radius of the plant. The transportation facilities and topographic feat ures of the LPZ are shown in Figure 2.1-7.

2.1.3.5 Population Center

The nearest population center is the City of Richland, 12 miles to the south.

2.1.3.6 Population Density In 2000, the population densities within the 10, 20, and 30-mile radii were 9, 96, and 73 people per square miles, respectively. In 2030, the dens ities out to the sa me distances are estimated to be 13, 123, and 84, respectively, based on the projections in Table 2.1-1

.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 2.1-7 2.

1.4 REFERENCES

2.1-1 Letter from Atomic Energy Commiss ion, Richland Operations Office, to Managing Director of the Supply Syst em, Washington Public Power Supply System,

Subject:

Appendix 2P, November 25, 1970.

2.1-2 Yandon, K. E., Projections and Distributions of Popul ations Within a 50-Mile Radius of Washington Public Power Supply System Nuclear Projects Nos. 1, 2, and 4 by Compass Direction and Rad ii Intervals, 1970-2030, October 1980.

2.1-3 Bonneville Power Admini stration, U.S. Department of Energy, Washington,

Subject:

Population, Employment a nd Household Project ions to 2000 by County, July 1979.

2.1-4 Bonneville Power Admi nistration, U.S. Depart ment of Energy, Oregon Population, Employment and Household Projections to 2000 by County.

2.1-5 Washington State Office of Financial Management, Forecasting Division (2012).

Small Area Estim ate Program: Census Block Groups

[SAEP_ColGenFac_Sectors_2000-2012.xl s]. Via communication with M. Mohrman, Washington Office of Fina ncial Management, July 2013; Oregon Population Data Estimates 2010 provid ed by Portland State University, Population Research Center, C. Ryne rson, in collaboration with the M. Mohrman at Washington State Office of Financial Manage ment, March 2014.

Table 2.1-1 Projected Population D i stribution by Comp a ss Sector and Distance from the Site 19 8 0 19 9 0 20 0 0 20 1 0 20 2 0 20 3 0 Distance Cumulative (mile s) Direction (comp a s s segm e n t) Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 LDC N-0 3-0 2 3 2.1-9 0-3 All 0 0 0 0 0 0 0 0 0 0 0 0 3-5 N-N N E 0 0 0 0 0 0 0 0 0 0 0 0 NE 10 10 35 35 48 48 52 52 55 55 86 86 E N E 22 32 43 78 56 104 60 112 63 118 64 150 E 22 54 43 121 56 160 60 172 63 181 64 214 ESE 22 76 43 164 56 216 60 232 63 244 64 278 SE 4 80 6 170 9 225 11 243 11 255 12 290 SSE-N N W 0 80 0 170 0 225 0 243 0 255 0 290 5-10 N 26 106 58 228 77 302 83 326 87 342 88 378 NNE 83 189 126 354 152 454 162 488 170 512 172 550

NE 155 344 198 552 224 678 240 728 252 764 254 804

ENE 114 458 157 709 177 855 190 918 200 964 202 10 0 6 E 135 593 200 909 257 11 1 2 276 11 9 4 290 12 5 4 293 12 9 9 ESE 168 761 276 11 8 5 341 14 5 3 366 15 6 0 385 16 3 9 389 16 8 8 SE 190 951 406 15 9 1 536 19 8 9 575 21 3 5 604 22 4 3 610 22 9 8 SSE 45 996 253 18 4 4 308 22 9 7 330 24 6 5 347 25 9 0 350 26 4 8 S 50 10 4 6 272 21 1 6 483 27 8 0 518 29 8 3 544 31 3 4 550 31 9 8 SSW 235 12 8 1 535 26 5 1 809 35 8 9 867 38 5 0 911 40 4 5 920 41 1 8 SW 25 13 0 6 25 26 7 6 25 36 1 4 27 38 7 7 28 40 7 3 29 41 4 7 WSW-NNW 0 13 0 6 0 26 7 6 0 36 1 4 0 38 7 7 0 40 7 3 0 41 4 7 10-2 0 N 332 16 3 8 371 30 4 7 398 40 1 12 427 43 0 4 449 45 2 2 454 46 0 1 NNE 328 19 6 6 371 34 1 8 397 44 0 9 426 47 3 0 447 49 6 9 452 50 5 3 NE 399 23 6 5 562 39 8 0 588 49 9 7 630 43 6 0 662 56 3 1 669 57 2 2 E N E 792 31 5 7 835 48 1 5 855 58 5 2 917 62 7 7 964 65 9 5 974 66 9 6 E 461 36 1 8 479 52 9 4 544 63 9 6 583 68 6 0 613 72 0 8 619 73 1 5 ESE 192 38 1 0 430 57 2 4 576 69 7 2 618 74 7 8 650 78 5 8 657 79 7 2 SE 41 5 5 79 6 5 52 2 1 10 9 45 58 2 1 12 7 93 62 4 2 13 7 20 65 6 1 14 4 19 66 2 7 14 5 99 SSE 49178 57143 63483 74428 70917 83710 76043 89763 79932 94351 80 7 34 95 3 33 S 28 9 43 86 0 86 37 6 72 11 2 100 45 4 34 12 9 144 48 7 17 13 8 480 51 2 08 14 5 559 51 7 22 14 7 055 SSW 15 9 2 87 6 78 17 7 2 11 3 872 19 2 2 13 1 066 20 6 1 14 0 541 21 6 6 14 7 725 21 8 8 14 9 243 SW 31 0 6 90 7 84 35 9 7 11 7 469 894 13 4 960 41 7 5 14 4 716 43 8 9 15 2 114 44 3 3 15 3 676 WSW 950 91 7 34 10 4 8 11 8 517 11 0 8 13 6 068 11 8 8 14 5 904 12 4 8 15 3 362 12 6 0 15 4 936 W 0 91 7 34 0 11 8 517 0 13 6 068 0 14 5 904 0 15 3 362 0 15 4 936 WNW 0 91 7 34 0 11 8 517 0 13 6 068 0 14 5 904 0 15 3 362 0 15 4 936 NW 0 91 7 34 0 11 8 517 0 13 6 068 0 14 5 904 0 15 3 362 0 15 4 936 NNW 0 91 7 34 0 11 8 517 0 13 6 068 0 14 5 904 0 15 3 362 0 15 4 936 Table 2.1-1 Projected Population D i stribution by Compass Sec t or and Distance from the Site (Continued) 19 8 0 19 9 0 20 0 0 20 1 0 20 2 0 20 3 0 Distance Cumulative (mile s) Direction (comp a s s segm e n t) Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 LDC N-0 3-0 2 3 2.1-10 20-3 0 N 15 0 1 93 2 35 18 3 7 12 0 354 20 5 5 13 8 123 22 0 3 14 8 107 23 1 6 15 5 678 23 3 9 15 7 275 NNE 57 5 9 98 9 94 64 8 7 12 6 841 71 2 3 14 5 246 76 3 8 15 5 745 80 2 9 16 3 707 81 1 0 16 5 385 NE 20 1 5 10 1 009 21 7 4 12 9 015 22 7 4 14 7 520 24 3 8 15 8 183 25 6 3 16 6 270 25 8 9 16 7 974 E N E 17 1 7 10 2 726 17 6 0 13 0 775 17 8 6 14 9 306 19 1 5 16 0 098 20 1 3 16 8 283 20 3 3 17 0 007 E 151 10 2 877 194 13 0 969 220 14 9 526 236 16 0 334 248 16 8 531 250 17 0 257 ESE 153 10 3 030 240 13 1 209 305 14 9 831 327 16 0 661 344 16 8 875 348 17 0 605 SE 61 3 8 10 9 168 65 1 2 13 7 721 67 3 8 15 6 569 72 2 5 16 7 886 75 9 4 17 6 469 76 7 0 17 8 275 SSE 24 1 16 13 3 284 32 5 59 17 0 280 36 3 60 19 2 929 38 9 87 20 6 873 42 0 32 21 8 501 42 4 54 22 0 729 S 187 13 3 471 678 17 0 958 975 19 3 904 10 4 5 20 7 918 10 9 8 21 9 599 11 0 9 22 1 838 SSW 875 13 4 346 12 1 8 17 2 176 14 2 6 19 5 330 15 2 9 20 9 447 16 0 7 22 1 206 16 2 3 22 3 461 SW 61 6 5 14 0 511 71 4 7 17 9 323 77 3 7 20 3 067 82 9 6 21 7 743 87 2 0 22 9 926 88 0 8 23 2 269 WSW 16 2 6 14 2 137 17 9 9 18 1 122 19 0 8 20 4 975 20 4 6 21 9 789 21 5 1 23 2 077 21 7 3 23 4 442 W 11 9 1 14 3 328 13 2 5 18 2 447 14 2 9 20 6 404 15 3 2 22 1 321 16 1 0 23 3 687 16 2 6 23 6 068 WNW 185 14 3 513 280 18 2 727 297 20 6 701 318 22 1 639 334 23 4 021 338 23 6 406 NW 40 14 3 553 44 18 2 771 48 20 6 749 51 22 1 690 54 23 4 075 55 23 6 461 NNW 182 14 3 735 200 18 2 971 218 20 6 967 234 22 1 924 246 23 4 321 249 23 6 710 30-4 0 N 980 14 4 715 10 9 6 18 4 065 11 2 7 20 8 094 12 0 8 22 3 132 12 7 0 23 5 591 12 8 3 23 7 993 NNE 31 9 8 14 7 913 36 6 3 18 7 728 39 8 3 21 2 077 42 7 1 22 7 403 44 9 0 24 0 081 45 3 6 24 2 529 NE 650 14 8 563 800 18 8 528 745 21 2 822 799 22 8 202 846 24 0 927 850 24 3 379 E N E 421 14 8 984 447 18 8 975 475 21 3 297 509 22 8 711 535 24 1 462 540 24 3 919 E 128 14 9 112 136 18 9 111 141 21 3 438 152 22 8 863 160 24 1 622 162 24 4 081 ESE 167 14 9 279 176 18 9 287 182 21 3 620 195 22 9 058 205 24 1 827 208 24 4 289 SE 464 14 9 743 484 18 9 771 497 21 4 117 533 22 9 591 560 24 2 387 566 24 4 855 SSE 592 15 0 335 844 19 0 615 955 21 5 072 10 2 3 23 0 615 10 7 6 24 3 463 10 8 7 24 5 942 S 46 8 0 15 5 015 56 5 3 19 6 268 63 6 8 22 1 440 68 2 8 23 7 442 71 7 2 25 0 635 72 5 0 25 3 192 SSW 256 15 5 271 424 19 6 692 529 22 1 969 567 23 8 009 596 25 1 231 602 25 3 794 SW 473 15 5 744 661 19 7 353 786 22 2 755 842 23 8 851 885 25 2 116 894 25 4 688 WSW 21 8 71 17 7 615 24 7 29 22 2 082 26 8 90 24 9 645 28 8 33 26 7 684 30 3 62 28 2 478 30 6 65 28 5 353 W 35 7 8 18 1 193 39 4 9 22 6 031 42 7 3 25 3 918 45 8 2 27 2 266 48 1 6 28 7 294 48 6 4 29 0 217 WNW 13 9 9 18 2 592 14 5 9 22 7 490 15 7 9 25 5 497 16 9 3 27 3 959 17 8 0 28 9 074 17 9 8 29 2 015 NW 703 18 3 295 770 22 8 260 836 25 6 333 896 27 4 855 942 29 0 016 952 29 2 967 NNW 15 7 5 18 4 870 17 3 8 22 9 998 18 9 9 25 8 232 20 3 6 27 6 891 21 4 0 29 2 156 21 6 1 29 5 128 40-5 0 N 17 8 72 20 2 742 19 7 30 24 9 728 21 5 72 27 9 804 23 1 30 30 0 021 24 3 12 31 6 468 24 5 56 31 9 684 NNE 893 20 3 635 10 1 9 25 0 747 11 2 1 28 0 925 12 0 2 30 1 223 12 6 3 31 7 731 12 7 5 32 0 959 NE 926 20 4 561 11 3 9 25 1 886 12 7 5 28 2 200 13 6 7 30 2 590 14 3 7 31 9 168 14 5 1 32 2 410 E N E 213 20 4 774 243 25 2 129 375 28 2 575 402 30 2 992 423 31 9 591 427 32 2 837 E 241 20 5 015 258 25 2 387 268 28 2 843 287 30 3 279 302 31 9 893 305 32 3 142 ESE 864 20 5 879 925 25 3 312 961 28 3 804 10 3 0 30 4 309 10 8 3 32 0 976 10 9 5 32 4 237 Table 2.1-1 Projected Population D i stribution by Compass Sec t or and Distance from the Site (Continued) 19 8 0 19 9 0 20 0 0 20 1 0 20 2 0 20 3 0 Distance Cumulative (mile s) Direction (comp a s s segm e n t) Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total Numb er Cumulative Total C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT Dece m ber 2003 LDC N-0 3-0 2 3 2.1-11 40-50 (c ont.) SE 20 8 4 20 7 963 22 4 5 25 5 57 23 4 9 28 6 153 25 1 8 30 6 827 26 4 6 32 3 622 26 7 3 32 6 910 SSE 17 4 0 20 9 703 19 2 0 25 7 477 20 7 2 28 8 225 22 2 2 30 9 049 23 3 6 32 5 958 23 5 9 32 9 269 S 16 5 40 22 6 243 16 4 06 27 3 883 17 7 08 30 5 933 18 9 87 32 8 036 19 9 58 34 5 916 20 1 58 34 9 427 SSW 26 1 0 22 8 853 28 9 5 27 6 778 29 7 2 30 8 905 31 8 6 33 1 222 33 4 9 34 9 265 34 2 8 35 2 855 SW 421 22 9 274 443 27 7 221 476 30 9 381 509 33 1 731 535 34 9 800 541 35 3 396 WSW 809 23 0 083 892 27 8 113 965 31 0 346 10 3 5 33 2 766 10 8 8 35 0 888 10 9 9 35 4 495 W 18 5 15 24 8 598 20 4 81 29 8 594 22 1 76 33 2 525 23 7 80 35 6 546 24 9 96 37 5 884 25 2 47 37 9 742 WNW 17 4 2 25 0 340 19 0 3 30 0 497 20 4 3 33 4 568 21 9 1 35 8 737 23 0 3 37 8 187 23 2 6 38 2 068 NW 812 25 1 152 859 30 1 356 905 33 5 473 970 35 9 707 10 2 0 37 9 207 10 3 0 38 3 098 NNW 532 25 1 684 587 30 1 943 642 33 6 115 688 36 0 395 723 37 9 930 730 38 3 828

Table 2.1-2 2010 Population Distributi on by Compass Sector and Distance from the Site LDCN-14-009,13-067 2.1-12 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Distance (miles) Direction (compass segment) 2010 Population Distance (miles) Direction (compass segment) 2010 Population Distance (miles) Direction (compass segment) 2010 Population 0-3 ALL 0 5-10 N 16 10-20 N 136 5-10 NNE 78 10-20 NNE 742 3-4 NNE 0 5-10 NE 248 10-20 NE 1699 3-4 ENE 0 5-10 ENE 213 10-20 ENE 934 3-4 E 1 5-10 E 298 10-20 E 680 3-4 ESE 0 5-10 ESE 308 10-20 ESE 513 3-4 SE-NNW 0 5-10 SE 588 10-20 SE 16219 5-10 SSE 239 10-20 SSE 95379 4-5 NNE 1 5-10 S 376 10-20 S 38669 4-5 NE 18 5-10 SSW 548 10-20 SSW 7516 4-5 ENE 19 5-10 SW 16 10-20 SW 1232 4-5 E 27 5-10 WSW-NW 0 10-20 WSW 11 4-5 ESE 25 5-10 NNW 1 10-20 WNW 0 4-5 SE 0 10-20 NNW 1 4-5 SSE-NNW 0 10-20 NW - W 0 0-5 TOTAL 91 0-10 TOTAL 3020 0-20 TOTAL 166751 20-30 N 1098 30-40 N 1138 40-50 N 37171 20-30 NNE 13285 30-40 NNE 3802 40-50 NNE 782 20-30 NE 1237 30-40 NE 210 40-50 NE 705 20-30 ENE 3638 30-40 ENE 302 40-50 ENE 164 20-30 E 53 30-40 E 121 40-50 E 48 20-30 ESE 509 30-40 ESE 686 40-50 ESE 188 20-30 SE 18611 30-40 SE 343 40-50 SE 1415 20-30 SSE 50050 30-40 SSE 201 40-50 SSE 310 20-30 S 1710 30-40 S 8330 40-50 S 30053 20-30 SSW 131 30-40 SSW 236 40-50 SSW 5441 20-30 SW 11225 30-40 SW 1552 40-50 SW 217 20-30 WSW 1475 30-40 WSW 41107 40-50 WSW 4428 20-30 W 50 30-40 W 934 40-50 W 22691 20-30 WNW 205 30-40 WNW 6111 40-50 WNW 26 20-30 NW 611 30-40 NW 2478 40-50 NW 619 20-30 NNW 443 30-40 NNW 3870 40-50 NNW 1622 0-30 TOTAL 271082 0-40 TOTAL 342503 0-50 TOTAL 448383 Table based on April 2010 Ce nsus Bureau counts.

Sandpit Eltopia Ranch Water Tank 17 17 395 Hathaway 800Basin City 770 170 Ranch Savage IslandBasin City Ranch Ranch Frischnecht Shano Radio Tower Radio Tower Mesa SMITH CANYON OLD MAID Water Tank Othello Air Force Sta CactusSaddle Gap Potholes East Canal US Government RR Mathews Corner RINGOLD FLAT 260 Camp Lake Scooteney Reservoir T Lake WASHTUCNA COULEE Clark Pond P A R A D I S E F L A T S Canal Eagle Lakes SURVIVAL TRAINING AREA JUNIPER FOREST Water Tank Gravel pit Gravel pit Radio Tower Radio Tower Humorist 12 Quarry RYE GRASS COULEE Water Tank 124 Gravel pit Glade Pasco Substa Substa Ice Harbor Dam JACKASS MOUNTAIN Water Tank Wooded Island ESQUATZEL COULEE Diversion Channel Esquatzel RichlandPasco West Vista 521 Columbia Point Tri-Cities Airport 407 Potholes Canal WT Pipeline 395 Sacajawea State Park SNAKE RIVER UN ION PAC IFIC Hedges WT Substa 14 Vista Columbia Canal KENNEWICK Franklin Co Irr Can NORTHERN Columbia Canal BADGER MOUNTAIN Canal Island View Richland 391 221 Grain elevator Grain elevator Gravel pit 240 224 Ranch Radio TowerGrain elevatorGrain elevator Badger Canyon BadgerCandy Mtn.

Fast Flux Test Facility RATTLESNAKE HILLS Water Tank Water TanksKelly Gulch Horn Rapids Dam WTCorral Canyon WT Radio Tower Observatory Pumping Station McWhorter 1356 West Richland RED MOUNTAINGoose Hill UNION PACIFIC Benton City BURLINGTON Gibbon WhitstrandChandler Butte Sunnyside Canal Chaffee Pumping Station Ranch RanchRoza Canal UNION PACIFIC Prosser SpringWebber Canyon Roza CanalBlack Canyon Spring CrSnipes Creek 24 243 24 W A H L U K E S L O P E Ranch Christensen Bros 840 Mattawa 750White Bluffs SADDLE MOUNTAIN NATIONAL WILDLIFE REFUGE CABLE BUTTE Hanford CABLE MOUNTAINCoyote RapidsHanford Ditch Towers Water Tank Benson Spring Locke Island Substa BENTON CO FRANKLIN CO 82 Dam North Prosser 82 12 12 82 Kiona Gravel pit 182 182 12 Chandler Island View Sacajawea State Park Route 4 South Route 4 South Route 11A Basin Hill Rd.Hollingsworth Rd.Bellflower Rd.Columbia Rd.Mountain Vista RdRoute 10 South Route 2 South Horn Rapids Rd.Stevens Dr.Selph Landing Rd.

Taylor Flats Rd.

Glade North Rd.

Alder Rd Birch Rd.W. Sagemoor Rd.

Cedar Dogwood Elm Rd.West Fir Rd.Eltopia Rd.

Fir Rd.Fir Way Glenwood Glenwood Auburn Rd.

N. Belleview Rd.N. Belleview Rd.N. Coulee Rd.

CootonIronwood Rd.

W. Juniper Rd.Russell Rd.

W. Klamath Rd.

Sheffield Rd.Juniper Rd.Olympia Dr.

Glade North Rd.

Ash R 170 Harrington VanGiesonGeorge Washinton Way N. Rd 68 W. Sagemoor Rd.

Finley Prosser 694 MCNARY NATIONAL WILDLIFE REFUGE WALLA Stevens Country Christian Country Haven Edwin Markum Big River COLD CREEK VALLEY H A N F O R D R E S E R V A T I O N YAKIMA RIVER Iowa Flats N Power Plant Loop Columbia Generating Station Kinder-Care KiBe Burbank Hi wood GRANT CO20 Statute Miles 15 10 5 0 530 Kilometers 25 15 10 5 0 5 20 SCALE Railroad Pipeline PowerlineMedium Duty Road Light Duty RoadHeavy Duty Road 395 260 Landplane Airport Landing AreaFederal Route MarkerState Route

Interstate Route 80NPark or Reservation BoundarySandpit or Gravelpit Populated Places Radio Tower, Well Latitude/LongitudeEmergency Center LEGEND 5 MILES10 MILES20 MILES FFTF Hepner JunctionPin e C reek pit pit Sandpit Eltopia Ranch Water Tank Water Tank 17 260 260 26 261 260 Gravel pit 17 Five Corners 395 395 Hathaway 800 Pent 884 Watson 1349Baumann Farm 1600 Othello 1145 Basin City 770 170 Ranch Savage Island Basin City Ranch Ranch Lower Monumental DamPleasant View Farm FIELDS GULCH Ayer Ranch PALOUSE FALLS RECREATION AREA Farm Harder DUNNIGAN COULEE Davin Skookum Can 261 Grain Elevator Grain Elevator Old RR grade Connell Connell City 925 Grain Elevator WASHTUCNA COULEE Rattlesnake Can COULEE HARDESTY Michigan P ra irie Washtucna Kahlotus SAND HILLS COULEE Hatton HATTON C OULEE ENE E R A T T L E S N A K E F L A TCunningham Canyon PROVIDENCE COULEEPipeline Grain Elevator Cunningham Frischnecht Shano Bruce BURLINGTO NNORTHER N Radio Tower Radio Tower Mesa Devils Canyon SMITH CANYON OLD MAID COULE E Water Tank SITE Othello Air Force Sta Cactus OthelloEL E CT RIFIE D Taunton Saddle Gap Potholes East Canal US Governm ent RR Ranch BURLINGTON NORTHERN UNION PACIFIC Lower Monumental 813Farrington pit Scott Magallon Mathews Corner RINGOLD FLAT Fisher Ranch 1521 10 MI 20 MI 30 MI 40 MI 50 MI 260 Rattlesnake Lake Camp Lake Scooteney Reservoir T Lake WASHTUCNA COULEE Clark Pond Lake Kahlotus Sulfur Lake P A R A D I S E F L A T S Low Canal Eagle Lakes Dry Lake ESE Prescott Harsha Lamar Eurica Grain elevator Grain elevator ElwoodClyde Grain elevator Hatch Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Hadley Walker Page Valley Grove Ennis Grain elevator Grain elevator Grain elevator Whitman Station Substa Substa College Place WALLA WALLA Lowden Touchet Martin Slater Pipeline 12 395 Zanger Junction Reese Climax Siding Grain elevator SURVIVAL TRAINING AREA JUNIPER FOREST Water Tank Gravel pit Gravel pit Redd Levey Water Tank Grain elevator Grain elevator JOHNSON BUTTE The Butte Radio Tower Radio Tower Wallula Water Tank Humorist MCNARY NATIONAL WILDLIFE REFUGE Nine Mile Canyon Compressor Station 540 Snake River 12 Quarry Walker Canyon RYE GRASS FLATEEL U OC SSARG EYR Water Tank Quarry 124 Lake Sacajawea Gravel pit Glade Pasco Substa Substa Ice Harbor Dam JACKASS MOUNTAIN Water Tank Wooded Island Taylor Flat ESQUATZEL COULEE Diversion Channel Esquatzel Richland Pasco West Vista 521 Columbia Point Tri-Cities Airport 407 Potholes Canal WTPipeline Dry Hollow E U R E K A F L A T Winnett Canyon Winnett Canyon Badger Hollow Badger Hollow Gravel pitBURLINGTON NORTH ERNBUR LINGTO N N ORT HER N SPRING VALLEYWebber Canyon W oodward C anyon Burbank Finley Nine Canyon 395 Sacajawea State Park WT SNAKE RIVER L A K E W A L L U L A UNION PACIFICUNION PA CIFIC Hedges WT Substa Grain elevator Gravel pit Radio Towers Radio Tower Grain elevator Williams Wells Amon Well Young Wells Coyote Spring 14 Bowden Springs Bofer Canyon Taylor Canyon Four Canyon Straub Canyon Vista Cress Wells Columbia Canal KENNEWICKFranklin Co Irr Can N ORTHERNBURLING T O N NO RTHERN C olumbia CanalBAD G E R M O U N TA IN Canal Switzler CanyonPipeline HORSE HEAVEN HILLS D ry Creek Walla Walla River WHITMAN MISSIONNHSWalla Walla R iver Island View Richland 391 124 260 McNary Dam L A K E W A L L U L A WASHINGTON OREGON SE S Gardena Windmill Big Well Cold Springs Reservoir Hunt DitchGreasewood Creek Pine Dry Cr Burling a m e D itc h To P endleton 395 730 730 30 84 Umatilla River 207 32 SSE Milton-Freewater Walla Walla Valley Umapine Walla Walla River Skyranch 850 Page 850 204 Weston Athena 11 11 Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator WT WT Bowlus Hill Grain elevator Oregon Sky Ranch 357 Grain elevator Holdman WT NORTH FORK JUNIPER CANYON NORTH FORK COLD SPRING S CANY O N SO UTH F O RK JUNIPER CAN YON MIDDLE FORK COLD SPRINGS CANYON SOUTH FORK COLD SPRINGS CANYON COLD SPRINGS CANYON DESPAIN GULCH S T A GE G ULCH JUNIPER CANYONRUSH CA NYON Switzler Canyon Tank BURLINGTON NORTHERN WARM SPRINGS CANYON VA NSYCLE C ANYO N Grain elevator 37 FOURMILE GAP Echo HAT ROCK STATE PARK UMATILLA BUTTE Hermiston State Stanfield Hinkle SILLUSI BUTTE HERMISTON BUTTE Hermiston McNary Power City Substation HERMISTON BUTTEUNION PACIFIC 207 COLD SPRINGS NWR 84 221 SSW 14 14 BOARDMAN BOMBING RANGE 730LAKE UMATILLAUMATILLA NATIONAL WILDLIFE REFUGE COLUMBIA RIVER Hepner Junction GOLGOTHA BUTTE 84 CANOE RIDGE ALDER RIDGE ALDER RIDGE Windmill Windmill Boardman 390 PATERSON DEAD CANYON TULE CANYON Rattlesnake Spring Ranch Ranch Ordnance Water tower Umatilla Army Depot RIDGE UMATILLA ORDNANCE DEPOT Westland West Extension Irrigation Canal Disposal Plant Irrigon UNION PACIFIC Water Tower BURLINGTON Paterson FOURMILE CANYON WARD BUTTE FINLEY BUTTES BING CANYON East Branch Glade Creek COYOTE COULEE Boardman Coyote Canyon SAND RIDGE Windmill Ranch Windmill Ranch Cemetery Cleveland Bickleton Cemetery STEGEMAN CANYON SW BIG HORN CANYON WOOD GULCH DOUTY CANYON Ald er C reek Pine Creek SPRING CANYON Crow Butte 221 Grain elevator Grain elevator Grain elevator Gravel pit 240 224 241 223 22 22 Ranch H O R S E H E A V E N H I L L S Cem Ranch Grain elevator Ranch Grain elevator Ranch Radio Tower Gravel pit Gravel pit Ranch Grain elevator Carter Canyon Grain elev Grain elevator Grain elevator Grain elevator Grain elevator Ranch Badger Canyon Badger Candy Mtn.

Fast Flux Test Facility RATTLESNAKE HILLS Water Tank Water Tanks Kelly Gulch Horn Rapids Dam WT Corral Canyon WT Radio Tower Observatory Pumping Station McWhorter 1356 West Richland Iowa FlatsRED M OUNTAIN Goose Hill UNION PACIFIC Benton City BURLINGTON Gibbon Whitstrand Chandler Butte Sunnyside Canal Chaffee Pumping Station Ranch Ranch Roza Canal YAKIM A RIVER UNION PACIFIC HORSE HEAVEN HILLS Prosser 694 GRANDVIEW BUTTE Pumping Station Sunnyside 764 Lichty Sunnyside Bryon Ranch Pumping Station Pumping Station Toppenish Buena Outlook Emerald Granger Giffen Lake TOPPENISH NWR WSW Yakima I R Hdqrs NO RT H ERN BUR LING T ON Granger Farm

Labor Camp Pumping Station TOPPENISH NWR SNIPES MOUNTAIN BURLIN GTON HEMBRE MTN Satus Mabton WT Cem Pumping Station Sulfur Spring Water Tanks Radio Tower Liberty Oleys Lake Grange Mabton West Canal Yakima Indian Reservation Grain elevator Bluelight Windmill Grain elevator Radio Station Radio Tower Moore Canyon Coyote Canyon Gravel pit Prosser pit 97 97 Lozier Springs Campbell Spring Maiden Spring Spring Webber Canyon Mabton Siphon Toppenish Creek Toppenish C reekWapity Slough Satus No. 2 Pump Canal Sunnyside Canal Zillah Roza Canal Sulfur Cr Black CanyonSpring Cr Snipes Creek Water 410 PipelineEast Branch Glade Cr McKinley Spring Glade Creek Glade C re ek WNW W 24 24 243 24 Ranch Ranch Range Central Cem Observation post Badger Gap Squaw Tit Firing Center AAF Y A K I M A R I D G E U M T A N U M R I D G E Y A K I M A F I R I N G C E N T E R SADD LE MOUNT A IN S SADDLE MOUNTAINS CORRAL CANYON SOURDOUGH CA NYO N ALKALI CANYON SADDLE MOUNTAINS NW N NE W A H L U K E S L O P E Ranch Priest Rapids Dam Mattawa Christensen Bros 840 SMYRNA BENCH CNWR CNWR CNWR Beverly Ranch WAHATIS PEAK Mattawa 750 Desert 570 Corfu Ranch ROYAL SLOPE Water Tank CNWR CNWR CNWR Schwana White Bluffs SADDLE MOUNTAIN NATIONAL WILDLIFE REFUGE CABLE BUTTE Hanford CABLE MOUNTAIN Coyote Rapids Lower Crab Creek WT CNWR Cold Creek Hanford Ditch Towers Ranch Ranch Water TankCOLD CREE K VA LLEY Benson Spring Locke Island Dry Cr Horsethief Pt Cairn Hope Peak Emerson Nipple Substa H A N F O R D R E S E R V A T I O N Wapato Sawyer Donald Pumping station ELEPHANT MOUNTAIN HILLS ZILLAH PEAK M OXEE VA LLEY Moxee City RATTLESNAKE HILLS Ranch Tower Plant Quarantine Station BLAC K ROCK VALLEY Ranch Ranch Cem High Top R A T T L E S N A K E H I L L S R A T T L E S N A K E H I L L S Radio Facility KITTITAS CANYONMoxee Canal Moxee Canal Selah Creek Hanson Creek McDonald Springs Cottonwood Creek Black Rock Spring Sentinel GapSentinel Bluffs ELECTRIFIED Smyrna SENTINEL MOUNTAIN Cem Priest Rapids Lake NNW SAND DUNES WHISKEY DICK MTN ROCKY COULEE S PRING GULCH Caribou Creek RYEGRASS COULEE PARK MIDDLE CANYON HULT BUTTEJOHNSON CANYON BOYLSTON MOUNTAINS Rock Spring Poison Spring Lone Star Spring RYEGRASS MOUNTAIN Wippe Pumping Station Badger C r Fire station Kittitas Microwave Tower SCHNEBLY COULEE F R E N C H M A N H I L L S Winchester Wasteway Moses Lake Sky Ranch 1100 George SAND HILLS Potholes State Park THE POTHOLES 10 Mae 10 Ranch 90 90 26 CNWR Christensen 1150 F R E N C H M A N H I L L S Cem Camp Royal City NATURAL CORRAL LARSON AFB IMPACT AREA Red Rock Coulee CNWR W est Canal Frenchman Hills Lake Royal Lake Wanapum Dam CNWR Frenchman Springs POTHOLES COULEE BABCOCK BENCH LAVA GINKGO PETRIFIED FOREST PARK Sand Hollow BABCOCK RIDGE Stan Coffin Lake CRESCENT BAR CRESCENT BAR 281 283 Leslie Creek Skookumchuck Creek Johnson Cr Boylston BADGER POCKET East Kittitas Airway BeaconTrail Creek North Branch Canal Beacon HIAWATHA VALLEY SAND DUNES 17 17 10 90 Larson Air Force Base 1186 Moses Lake POTHOLES RESERVOIR Moses Lake 1200 21 21 261 261 Ritzville Ritzville 1800 21 395 NNE Lind Dewald 1880 Warden Water Franz Ranch Schrag Station Laing Station Refinery Wheeler Raugust Station Windmill LIND COULEE Radio Tower Dryland Experiment Providence Grain elevators Grain elevators Lind 1491 Servia LIND COULEE BOWERS COULEE LIND CO ULEE Pipeline New Warden 1265 Grain elevator Gravel pit Pipeline NORTHERN PACIFIC Underground aqueduct CHICAGO MILWAUKEE ST PAUL AND PACIFIC FARRIER COULEE Windmill WindmillPAHA COU LEE Paha NORTHERN PACIFIC 10 90 Ralston McElroy Lake Pizarro Grain elev ROCKY COULEE Underground aqueduct Lewis Horn Pelican Horn LIND CO ULEE Bassett Junction McDonald Siding Windmill Tiflis Ritell Siding NORTHERN PACIFICCHICAGO MILWAUKEE S T PAUL AND PACIFIC Rocky Coule e WastewayO'Sullivan Dam Sielen Siding Upper Goose Lake JACKASS MOUNTAIN Parker Horn Soda Lake COLUMBIA NATIONAL WILDLIFE REFUGE Goose Lake Shiner Lake Morgan Lake Taggars 1149 Kent Farms 1155 Beatrice Gravel pit Roxboro WEBER COULEE Meter Station Hillcrest Ruff KITTITAS CO GRANT CO ADAMS CO BENTON CO FRANKLIN CO YAKIMA CO KLICKITAT CO WALLA WALLA CO UMATILLA CO MORROW CO 82 Plymouth 82 12 82 82 12 Cem Grandview Dam North Prosser 82BURLIN G TON 12 12 82 Kiona Gravel pit 182 182 12 Chandler Island View Sacajawea State Park Sixprong Creek Sudbury Route 4 South Route 4 South Route 11A Basin Hill Rd.

Hollingsworth Rd.

Bellflower Rd.

Columbia Rd.

Mountain Vista Rd Route 10 South Route 2 South Horn Rapids Rd.

Stevens Dr.

Selph Landing Rd.

Taylor Flats Rd.

Glade North Rd.

Alder Rd Birch Rd.W. Sagemoor Rd.

Cedar Dogwood Elm Rd.West Fir Rd.

Eltopia Rd.

Fir Rd.Fir Way Glenwood Glenwood Auburn Rd.

N. Belleview Rd.N. Belleview Rd.N. Coulee Rd.

Cootonwood Ironwood Rd.

W. Juniper Rd.

Russell Rd.

W. Klamath Rd.

Sheffield Rd.

Juniper Rd.

Olympia Dr.

Glade North Rd.

Ash R 170 Harrington VanGieson George Washinton Way N. Rd 68 W. Sagemoor Rd.

46°00'46°30'46°45'46°15'45°45'47°00'Longitude Latitude 46°00'46°30'46°45'46°15'47°00'Latitude Longitude 119°00'118°45'118°30'119°30'120°00'119°45'119°15'120°15'119°00'118°45'118°30'119°30'120°00'119°45'119°15'120°15'12 395 395 395 395 30 3060 Miles50 Miles Pendleton 1493 37 Grain elevator ST AGE GULCH MISSOURI GULCH Railroad Pipeline Powerline Medium Duty Road Light Duty Road Heavy Duty Road LEGEND 395 260 Landplane Airport Landing Area Federal Route Marker State Route Interstate Route 80N Park or Reservation Boundary Sandpit or Gravelpit Populated Places Radio Tower, Well Latitude/Longitude 930114.50 Mi Aug 1998 SCALE 20 Statute Miles 15 10 5 0 5 30 Kilometers 25 15 10 5 0 5 20 45°45'Brushy Creek Eltopia Ranch Water TankBasin City 770 170 Ranch Savage IslandBasin City Ranch Ranch US Government RR Mathews Corner RINGOLD FLAT Camp Lake Clark Pond Radio Tower Gravel pit Glade Pasco JACKASS MOUNTAIN Water Tank Wooded Island ESQUATZEL COULEE Diversion Channel Esquatzel RichlandPasco West Vista 521 Columbia Point Tri-Cities Airport 407 Potholes Canal WT 395 Sacajawea State Park WT Substa 14 Vista KENNEWICK Franklin Co Irr Can NORTHERN Columbia Canal BADGER MOUNTAIN Island View Richland 391 Grain elevator 240 224Grain elevatorGrain elevator BadgerCandy Mtn.

Fast Flux Test Facility RATTLESNAKE HILLS Water Tank Water Tanks Horn Rapids

Dam WTCorral Canyon WT Radio Tower Observatory Pumping Station McWhorter 1356 West Richland RED MOUNTAINGoose Hill UNION PACIFIC Benton City BURLINGTONChandler Butte ChaffeeRoza CanalWebber CanyonWhite Bluffs CABLE BUTTE Hanford CABLE MOUNTAINCoyote RapidsHanford Ditch Towers BENTON CO 82 12 12 82 Kiona Gravel pit 182 182 12 Chandler Island View Sacajawea State Park Route 4 South Route 4 South Route 11A Basin Hill Rd.Hollingsworth Rd.Bellflower Rd.

Columbia Rd.Mountain Vista RdRoute 10 South Route 2 South Horn Rapids Rd.Stevens Dr.Selph Landing Rd.

Taylor Flats Rd.

Glade North Rd.

Alder Rd Birch Rd.W. Sagemoor Rd.

Cedar Dogwood Elm Rd.West Fir Rd.Eltopia Rd.

Fir Rd.Fir Way Glenwood Glenwood Auburn Rd.

N. Belleview Rd.N. Belleview Rd.N. Coulee Rd.

Cooton Ironwood Rd.

W. Juniper Rd.Russell Rd.

W. Klamath Rd.

Sheffield Rd.

Juniper Rd.Olympia Dr.

Glade North Rd.

Ash R 170 Harrington VanGiesonGeorge Washinton Way N. Rd 68 W. Sagemoor Rd.

Stevens Country Christian Country Haven Edwin Markum Big River COLD CREEK VALLEY H A N F O R D R E S E R V A T I O N YAKIMA RIVER Iowa Flats N Power Plant Loop Columbia Generating Station Kinder-Care KiBe woodScale in Miles 10 5 0 5 5 MILES20 MILES FFTF Canal10 MILES Hepner JunctionPin e C reek pit pit Sandpit Eltopia Ranch Water Tank Water Tank 17 260 260 26 261 260 Gravel pit 17 Five Corners 395 395 Hathaway 800 Pent 884 Watson 1349Baumann Farm 1600 Othello 1145 Basin City 770 170 Ranch Savage Island Basin City Ranch Ranch Lower Monumental DamPleasant View Farm FIELDS GULCH Ayer Ranch PALOUSE FALLS RECREATION AREA Farm Harder DUNNIGAN COULEE Davin Skookum Can 261 Grain Elevator Grain Elevator Old RR grade Connell Connell City 925 Grain Elevator WASHTUCNA COULEE Rattlesnake Can COULEE HARDESTY Michigan P ra irie Washtucna Kahlotus SAND HILLS COULEE Hatton HATTON C OULEE ENE E R A T T L E S N A K E F L A TCunningham Canyon PROVIDENCE COULEEPipeline Grain Elevator Cunningham Frischnecht Shano Bruce BURLINGTO NNORTHER N Radio Tower Radio Tower Mesa Devils Canyon SMITH CANYON OLD MAID COULE E Water Tank Othello Air Force Sta Cactus OthelloEL E CT RIFIE D Taunton Saddle Gap Potholes East Canal US Governm ent RR Ranch BURLINGTON NORTHERN UNION PACIFIC Lower Monumental 813Farrington pit Scott Magallon Mathews Corner RINGOLD FLAT Fisher Ranch 1521 10 MI 20 MI 30 MI 40 MI 50 MI 260 Rattlesnake Lake Camp Lake Scooteney Reservoir T Lake WASHTUCNA COULEE Clark Pond Lake Kahlotus Sulfur Lake P A R A D I S E F L A T S Low Canal Eagle Lakes Dry Lake ESE Prescott Harsha Lamar Eurica Grain elevator Grain elevator ElwoodClyde Grain elevator Hatch Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Hadley Walker Page Valley Grove Ennis Grain elevator Grain elevator Grain elevator Whitman Station Substa Substa College Place WALLA WALLA Lowden Touchet Martin Slater Pipeline 12 395 Zanger Junction Reese Climax Siding Grain elevator SURVIVAL TRAINING AREA JUNIPER FOREST Water Tank Gravel pit Gravel pit Redd Levey Water Tank Grain elevator Grain elevator JOHNSON BUTTE The Butte Radio Tower Radio Tower Wallula Water Tank Humorist MCNARY NATIONAL WILDLIFE REFUGE Nine Mile Canyon Compressor Station 540 Snake River 12 Quarry Walker Canyon RYE GRASS FLATEEL U OC SSARG EYR Water Tank Quarry 124 Lake Sacajawea Gravel pit Glade Pasco Substa Substa Ice Harbor Dam JACKASS MOUNTAIN Water Tank Wooded Island Taylor Flat ESQUATZEL COULEE Diversion Channel Esquatzel Richland Pasco West Vista 521 Columbia Point Tri-Cities Airport 407 Potholes Canal WTPipeline Dry Hollow E U R E K A F L A T Winnett Canyon Winnett Canyon Badger Hollow Badger Hollow Gravel pitBURLINGTON NORTH ERNBUR LINGTO N N ORT HER N SPRING VALLEYWebber Canyon W oodward C anyon Burbank Finley Nine Canyon 395 Sacajawea State Park WT SNAKE RIVER L A K E W A L L U L A UNION PACIFICUNION PA CIFIC Hedges WT Substa Grain elevator Gravel pit Radio Towers Radio Tower Grain elevator Williams Wells Amon Well Young Wells Coyote Spring 14 Bowden Springs Bofer Canyon Taylor Canyon Four Canyon Straub Canyon Vista Cress Wells Columbia Canal KENNEWICKFranklin Co Irr Can N ORTHERNBURLING T O N NO RTHERN C olumbia CanalBAD G E R M O U N TA IN Canal Switzler CanyonPipeline HORSE HEAVEN HILLS D ry Creek Walla Walla River WHITMAN MISSIONNHSWalla Walla R iver Island View Richland 391 124 260 McNary Dam L A K E W A L L U L A WASHINGTON OREGON SE S Gardena Windmill Big Well Cold Springs Reservoir Hunt DitchGreasewood Creek Pine Dry Cr Burling a m e D itc h To P endleton 395 730 730 Pendleton 1493 30 84 Umatilla River 207 32 SSE 37 Milton-Freewater Walla Walla Valley Umapine Walla Walla River Skyranch 850 Page 850 204 Weston Athena 11 11 Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator Grain elevator WT WT Bowlus Hill Grain elevator Oregon Sky Ranch 357 Grain elevator Holdman WT NORTH FORK JUNIPER CANYON NORTH FORK COLD SPRING S CANY O N SO UTH F O RK JUNIPER CAN YON MIDDLE FORK COLD SPRINGS CANYON SOUTH FORK COLD SPRINGS CANYON COLD SPRINGS CANYON DESPAIN GULCH S T A GE G ULCH JUNIPER CANYONRUSH CA NYON Switzler Canyon Tank BURLINGTON NORTHERN WARM SPRINGS CANYON VA NSYCLE C ANYO N ST AGE GULCH MISSOURI GULCH Grain elevator 37 FOURMILE GAP Echo HAT ROCK STATE PARK UMATILLA BUTTE Hermiston State Stanfield Hinkle SILLUSI BUTTE HERMISTON BUTTE Hermiston McNary Power City Substation HERMISTON BUTTEUNION PACIFIC 207 COLD SPRINGS NWR 84 221 SSW 14 14 BOARDMAN BOMBING RANGE 730 SCALE 20 Statute Miles 15 10 5 0 5 30 Kilometers 25 15 10 5 0 5 20LAKE UMATILLAUMATILLA NATIONAL WILDLIFE REFUGE COLUMBIA RIVER Hepner Junction GOLGOTHA BUTTE 84 CANOE RIDGE ALDER RIDGE ALDER RIDGE Windmill Windmill Boardman 390 PATERSON DEAD CANYON TULE CANYON Rattlesnake Spring Ranch Ranch Ordnance Water tower Umatilla Army Depot RIDGE UMATILLA ORDNANCE DEPOT Westland West Extension Irrigation Canal Disposal Plant Irrigon UNION PACIFIC Water Tower BURLINGTON Paterson FOURMILE CANYON WARD BUTTE FINLEY BUTTES BING CANYON East Branch Glade Creek COYOTE COULEE Boardman Coyote Canyon SAND RIDGE Windmill Ranch Windmill Ranch Cemetery Cleveland Bickleton Cemetery STEGEMAN CANYON SW BIG HORN CANYON WOOD GULCH DOUTY CANYON Ald er C reek Pine Creek SPRING CANYON Crow Butte 221 Grain elevator Grain elevator Grain elevator Gravel pit 240 224 241 223 22 22 Ranch H O R S E H E A V E N H I L L S Cem Ranch Grain elevator Ranch Grain elevator Ranch Radio Tower Gravel pit Gravel pit Ranch Grain elevator Carter Canyon Grain elev Grain elevator Grain elevator Grain elevator Grain elevator Ranch Badger Canyon Badger Candy Mtn.

Fast Flux Test Facility RATTLESNAKE HILLS Water Tank Water Tanks Kelly Gulch Horn Rapids Dam WT Corral Canyon WT Radio Tower Observatory Pumping Station McWhorter 1356 West Richland Iowa FlatsRED M OUNTAIN Goose Hill UNION PACIFIC Benton City BURLINGTON Gibbon Whitstrand Chandler Butte Sunnyside Canal Chaffee Pumping Station Ranch Ranch Roza Canal YAKIM A RIVER UNION PACIFIC HORSE HEAVEN HILLS Prosser 694 GRANDVIEW BUTTE Pumping Station Sunnyside 764 Lichty Sunnyside Bryon Ranch Pumping Station Pumping Station Toppenish Buena Outlook Emerald Granger Giffen Lake TOPPENISH NWR WSW Yakima I R Hdqrs NO RT H ERN BUR LING T ON Granger Farm

Labor Camp Pumping Station TOPPENISH NWR SNIPES MOUNTAIN BURLIN GTON HEMBRE MTN Satus Mabton WT Cem Pumping Station Sulfur Spring Water Tanks Radio Tower Liberty Oleys Lake Grange Mabton West Canal Yakima Indian Reservation Grain elevator Bluelight Windmill Grain elevator Radio Station Radio Tower Moore Canyon Coyote Canyon Gravel pit Prosser pit 97 97 Lozier Springs Campbell Spring Maiden Spring Spring Webber Canyon Mabton Siphon Toppenish Creek Toppenish C reekWapity Slough Satus No. 2 Pump Canal Sunnyside Canal Zillah Roza Canal Sulfur Cr Black CanyonSpring Cr Snipes Creek Water 410 PipelineEast Branch Glade Cr McKinley Spring Glade Creek Glade C re ek WNW W 24 24 243 24 Ranch Ranch Range Central Cem Observation post Badger Gap Squaw Tit Firing Center AAF Y A K I M A R I D G E U M T A N U M R I D G E Y A K I M A F I R I N G C E N T E R SADD LE MOUNT A IN S SADDLE MOUNTAINS CORRAL CANYON SOURDOUGH CA NYO N ALKALI CANYON SADDLE MOUNTAINS NW N NE W A H L U K E S L O P E Ranch Priest Rapids Dam Mattawa Christensen Bros 840 SMYRNA BENCH CNWR CNWR CNWR Beverly Ranch WAHATIS PEAK Mattawa 750 Desert 570 Corfu Ranch ROYAL SLOPE Water Tank CNWR CNWR CNWR Schwana White Bluffs SADDLE MOUNTAIN NATIONAL WILDLIFE REFUGE CABLE BUTTE Hanford CABLE MOUNTAIN Coyote Rapids Lower Crab Creek WT CNWR Cold Creek Hanford Ditch Towers Ranch Ranch Water TankCOLD CREE K VA LLEY Benson Spring Locke Island Dry Cr Horsethief Pt Cairn Hope Peak Emerson Nipple Substa H A N F O R D R E S E R V A T I O N Wapato Sawyer Donald Pumping station ELEPHANT MOUNTAIN HILLS ZILLAH PEAK M OXEE VA LLEY Moxee City RATTLESNAKE HILLS Ranch Tower Plant Quarantine Station BLAC K ROCK VALLEY Ranch Ranch Cem High Top R A T T L E S N A K E H I L L S R A T T L E S N A K E H I L L S Radio Facility KITTITAS CANYONMoxee Canal Moxee Canal Selah Creek Hanson Creek McDonald Springs Cottonwood Creek Black Rock Spring Sentinel GapSentinel Bluffs ELECTRIFIED Smyrna SENTINEL MOUNTAIN Cem Priest Rapids Lake NNW SAND DUNES WHISKEY DICK MTN ROCKY COULEE S PRING GULCH Caribou Creek RYEGRASS COULEE PARK MIDDLE CANYON HULT BUTTEJOHNSON CANYON BOYLSTON MOUNTAINS Rock Spring Poison Spring Lone Star Spring RYEGRASS MOUNTAIN Wippe Pumping Station Badger C r Fire station Kittitas Microwave Tower SCHNEBLY COULEE F R E N C H M A N H I L L S Winchester Wasteway Moses Lake Sky Ranch 1100 George SAND HILLS Potholes State Park THE POTHOLES 10 Mae 10 Ranch 90 90 26 CNWR Christensen 1150 F R E N C H M A N H I L L S Cem Camp Royal City NATURAL CORRAL LARSON AFB IMPACT AREA Red Rock Coulee CNWR W est Canal Frenchman Hills Lake Royal Lake Wanapum Dam CNWR Frenchman Springs POTHOLES COULEE BABCOCK BENCH LAVA GINKGO PETRIFIED FOREST PARK Sand Hollow BABCOCK RIDGE Stan Coffin Lake CRESCENT BAR CRESCENT BAR 281 283 Brushy Creek Leslie Creek Skookumchuck Creek Johnson Cr Boylston BADGER POCKET East Kittitas Airway BeaconTrail Creek North Branch Canal Beacon HIAWATHA VALLEY SAND DUNES 17 17 10 90 Larson Air Force Base 1186 Moses Lake POTHOLES RESERVOIR Moses Lake 1200 21 21 261 261 Ritzville Ritzville 1800 21 395 NNE Lind Dewald 1880 Warden Water Franz Ranch Schrag Station Laing Station Refinery Wheeler Raugust Station Windmill LIND COULEE Radio Tower Dryland Experiment Providence Grain elevators Grain elevators Lind 1491 Servia LIND COULEE BOWERS COULEE LIND CO ULEE Pipeline New Warden 1265 Grain elevator Gravel pit Pipeline NORTHERN PACIFIC Underground aqueduct CHICAGO MILWAUKEE ST PAUL AND PACIFIC FARRIER COULEE Windmill WindmillPAHA COU LEE Paha NORTHERN PACIFIC 10 90 Ralston McElroy Lake Pizarro Grain elev ROCKY COULEE Underground aqueduct Lewis Horn Pelican Horn LIND CO ULEE Bassett Junction McDonald Siding Windmill Tiflis Ritell Siding NORTHERN PACIFICCHICAGO MILWAUKEE S T PAUL AND PACIFIC Rocky Coule e WastewayO'Sullivan Dam Sielen Siding Upper Goose Lake JACKASS MOUNTAIN Parker Horn Soda Lake COLUMBIA NATIONAL WILDLIFE REFUGE Goose Lake Shiner Lake Morgan Lake Taggars 1149 Kent Farms 1155 Beatrice Gravel pit Roxboro WEBER COULEE Meter Station Hillcrest Ruff KITTITAS CO GRANT CO ADAMS CO BENTON CO FRANKLIN CO YAKIMA CO KLICKITAT CO WALLA WALLA CO UMATILLA CO MORROW CO Railroad Pipeline Powerline Medium Duty Road Light Duty Road Heavy Duty Road LEGEND 395 260 Landplane Airport Landing Area Federal Route Marker State Route Interstate Route 80N Park or Reservation Boundary Sandpit or Gravelpit Populated Places Radio Tower, Well Latitude/Longitude 82 Plymouth 82 12 82 82 12 Cem Grandview Dam North Prosser 82BURLIN G TON 12 12 82 Kiona Gravel pit 182 182 12 Chandler Island View Sacajawea State Park Sixprong Creek Sudbury 930114.50 Mi Aug 1998 Route 4 South Route 4 South Route 11A Basin Hill Rd.

Hollingsworth Rd.

Bellflower Rd.

Columbia Rd.

Mountain Vista Rd Route 10 South Route 2 South Horn Rapids Rd.

Stevens Dr.

Selph Landing Rd.

Taylor Flats Rd.

Glade North Rd.

Alder Rd Birch Rd.W. Sagemoor Rd.

Cedar Dogwood Elm Rd.West Fir Rd.

Eltopia Rd.

Fir Rd.Fir Way Glenwood Glenwood Auburn Rd.

N. Belleview Rd.N. Belleview Rd.N. Coulee Rd.

Cootonwood Ironwood Rd.

W. Juniper Rd.

Russell Rd.

W. Klamath Rd.

Sheffield Rd.

Juniper Rd.

Olympia Dr.

Glade North Rd.

Ash R 170 Harrington VanGieson George Washinton Way N. Rd 68 W. Sagemoor Rd.

46°00'46°30'46°45'46°15'45°45'47°00'Longitude Latitude 46°00'46°30'46°45'46°15'45°45'47°00'Latitude Longitude 119°00'118°45'118°30'119°30'120°00'119°45'119°15'120°15'119°00'118°45'118°30'119°30'120°00'119°45'119°15'120°15'12 395 395 395 395 30 30Columbia Generating Station

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-08-035 2.2-1 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES

This section describes the industr ial, transportation, and military installations and operations in the vicinity of the site which may have a potential effect on the safe operation of Columbia Generating Station (CGS).

2.2.1 LOCATION

AND ROUTES

There are no military bases, missile sites, manufacturing plants, chemical plants, commercial chemical storage facilities, or ai rports within a 5-mile radius of the site. A security barrier completely surrounds the sta tion and its major supporting facilities to keep unauthorized vehicles a safe distance from critical structures.

According to the Richland Operations Offi ce of the Department of Energy (DOE) (Reference 2.2-1), there are no plans for petrochemical storage facilities, airports, oil and gas pipelines, or petrochemical tank farms on the Hanford Site. Plan s for modifications to or new radiological material treatment or storage facilities are discussed in Section 2.2.2.

As shown in Figure 2.1-3, the following facilities are located at or near the CGS site:

Energy Northwest Plant Engineering Center, H. J. Ashe Substation, DOE Fast Flux Test Facility (FFTF), WNP-1 and WNP-4 sites, DOE 618-11 (Wye) radioactive waste burial ground, Permanent meteorological tower, Independent Spent Fuel Storage Installation (ISFSI), and Hydrogen Storage and Supply Facility.

Other facilities that are located within a 5-mile radi us of the site include:

The Plant Support Facility/Emergency Opera tion Facility which is located 0.75 miles southwest of CGS on Ener gy Northwest property, The Benton Substation which is located 3 miles east-southeast of CGS on DOE property, The Laser Interferometer Gravitational-Wa ve Observatory (LIGO) which is located approximately 3.3 mile s west-southwest of CGS on DOE property, and

The DOE 618-10 (300 North) radioactive waste burial ground which is located approximately 3.5 miles sout h of CGS on DOE property.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDCN-01-055,03-023 2.2-2 Transportation needs of CGS can be met by exis ting barge, rail, and highway facilities.

Barges of up to 3000 tons cap acity can be accommodated on th e Columbia River within the Hanford Site. A barge unloading facility 9 miles south of the plant was used for delivery of large construction items for the DOE FFTF and Energy Northwest nuclear projects. These materials were transported by truck or rail to the construction sites from the Port of Benton landing.

The CGS site is serviced by a tw o-lane paved access road connected to Hanford Site Route 4, which is a paved four-lane major artery located 1.

6 miles west of the station. Route 4 is part of the DOE road system. The DOE-owned road system connects the area s of the Hanford Site with paved two-lane and four-lane primary ro ads, secondary gravel roads, and unimproved roads. State Highway 240 traverses the Hanford Site from the southeast to the northwest. The highway passes within about 7 miles of CGS in the southwest quadrant. The highway connects into State Highway 24, which goes west to Yakima, Washington, and across the Vernita Bridge on the Columbia River 22 miles to the northwest (see Figure 2.2-1

).

The Hanford Site (DOE) railroad system (see Figure 2.2-2) connects with commercial rail systems in Richland and Kennewick, Washington. Railroad operations th at pass through CGS property are restricted to only those trains that have been authorized by Energy Northwest Security. The rail line is physically blocked at the two points where the plant vehicle barrier crosses the tracks.

Heavy barge traffic north of the Port of Benton doc k is not feasible because the river channel is too shallow and the current is too swift. The environmental impact and economic cost of

constructing a new barge slip at some upstream location a nd channeling the river cannot presently be justified with the availability of land transporta tion between the Port of Benton facility and the Hanford Site.

Making the Columbia River navigable for barges from north of Richland to Wenatchee would result in barge traffic past the CGS site at River Mile 352. However, this situation would not likely occur. Locks or other lift facilities would have to be constructed at the Priest Rapids, Wanapum, and Rock Island Dams. Furthermore, in 2000, a presidential executive order created the Hanford Reach National Monument, protecting the 51-mile Hanford Reach of the Columbia River (Reference 2.2-2). The protected area includes a 1/4-mile-wide corridor on the west side of the river in the vicinity of CGS.

Airports, military facilities, low-level Federal airways, and airport instrument approaches in the vicinity of CGS ar e discussed in Section 3.5.1.6 and shown in Figure 2.2-3.

An explosives and ordinance test site operated by Pacific Northwest National Laboratory approximately 13 miles northwest of the site was abandoned in mid-1975 (Reference 2.2-3). Explosives for operations such as quarrying or seismic studies on the Hanford Site are brought to the blasting site as needed and unused quantities are removed.

Normally the only explosives C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-011 2.2-3 stored on the Hanford Site are small arms ammunition for use by the security patrols. A small arms firing range used for training by the DOE security patrol is located 8 miles due south of the plant (Reference 2.2-3). Another range, used by Energy No rthwest security personnel, is located 1.5 miles east-nor theast of CGS on Energy Northwest lease property.

2.

2.2 DESCRIPTION

S

2.2.2.1 Description of Facilities Energy Northwest's Plant Engineering Center is located west of the CGS turbine generator building as shown in Figure 1.2-1. It is a two-story, 100,000 ft 2 facility designed to house approximately 470 CGS pl ant staff personnel.

The H. J. Ashe Substation is located approxim ately 0.5 mile north of CGS and is operated by the Bonneville Power Administration as part of its transmission system.

The Energy Northwest permanent meteorological tower is located less than 0.5 mile west of the plant site. The tower is automated so that the only personnel at the tower are those required to make adjustments to the instruments or to perform repairs to the system. There are no permanent personnel at the facility.

The ISFSI is located immediately north-northwest of the plant. Confinement of all radioactive materials at the ISFSI is provided by the required use of NRC certified spent fuel storage casks listed in 10 CFR 72.214. The ISFSI storage cask system consists of an inner stainless steel multi-purpose canister (MPC) and an outer st orage overpack. The MPC contains the spent fuel. It is a welded pressure vessel with no bolted closure or mechanical seals. Primary closure welds are examined and leakage tested to ensure thei r integrity. The MPC redundant closures are designed to maintain confinement integrity during normal conditions of storage, and off-normal and postula ted accident conditions. The outer storage overpack is fabricated from concrete and structural st eel components that are classifi ed as important to safety.

A fully loaded spent fuel storag e cask weighs approximately 185 tons. The spen t fuel loaded storage casks are located within the Energy Northwest ISFSI protected area which is surrounded by a fence and topped with barbed wire. The ISFSI access gates are locked except when in use.

The Hydrogen Storage and Supply F acility is located 0.6 miles south-southeast of the plant site. The facility is part of a hydrogen water chemistr y system to prevent and mitigate intergranular stress corrosion cracking in reactor internal struct ures and piping welds. The facility consists of a fenced gravel yard with concrete pads constructed to accommodate a liquid hydrogen tank, nitrogen tank, gaseous hydrogen tubes, and all supporting piping and equipment necessary to supply CGS with gaseous hydrogen. The liquid and compressed gases are delivered to the facility by truck.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-038 2.2-4 Within the exclusion area radius of 1950 m is Energy Northwests 1250-MWe WNP-1 project.

Construction of the PWR plant, located 1500 m (4925 ft) east-southeast of CGS, was suspended in April 1982. In May 1994 the Energy Northwest Board of Directors voted to terminate WNP-1. The construction of twin unit WNP-4, located 1250 m (4100 ft)

east-northeast of CGS, was terminated in January 1982. These projects have a separate access road that ties into the Hanford Site Route 4, 1.6 miles south of the CGS access road. Support

activities at either the WNP-1 or WNP-4 sites do not interfere with operation of CGS. These

may include activities associated with site restoration and economic development (such as

leasing of excess facilities for office space and manufacturing). Within the exclusion area, Energy Northwest has the authority to determine all activities within the meaning of 10 CFR 100.3(a), including the authority to remove all personnel and property from the area.

The Laser Interferometer Gravitational-Wave Observatory (LIGO) is located approximately 3.3 miles west-southwest from the plant site. The mission of this research facility is to observe gravitational waves of cosmic origin. The facility houses laser interferometers, consisting of mirrors suspended at each of the corners of an L-shaped vacuum system measuring 2.5 miles on a side. The materials and activities at this facility do not impact the operation of CGS.

As discussed in Section 2.1.1 and shown in Figure 2.1-1 , CGS is located on the DOE Hanford Site. In reviewing the plant site and the vicinity for potential external hazards or hazardous material, the Hanford facilities currently operating, recently operating, or with the potential for operating were screened. The facilities discussed below are those believed to pose the most risk to the safe operations of CGS. The safety analysis reports and accident analysis prepared

for those facilities were reviewed to determine possible hazards. No accidents evaluated present a physical challenge to the CGS buildings. Releases with the potential to impact the operation of CGS were radioactive particulate that would be effectively mitigated within

General Design Criterion (GDC) 19 limits by the control room high-efficiency particulate air (HEPA) filters. Considered but not included were the 200 East Burial Grounds, the Critical

Mass Laboratory, the Liquid Effluent Retention Facility, and the Effluent Treatment Facility in

the 200 East Area. In the 200 West Area, the T Plant, U Plant, Reduction-Oxidation Plant, and the 222-S Laboratory were considered but not included. These facilities have insufficient radiological or toxicological inventories in a dispersible form to represent a risk to CGS

operation. The specific facilities included are discussed in Table 2.2-1

.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-038 2.2-5 Three DOE facilities are located within a 5-mile radius of the plant site. These are the Fast

Flux Test Facility (FFTF) and two radioactive waste burial grounds. The specific hazards

associated with these facilities are summarized in Table 2.2-1 and the specific activities are listed below:

The FFTF is a deactivated sodium cooled breeder reactor located approximately 3 miles southwest of CGS. All fuel has been removed and shipped to the Idaho National

Laboratory. All sodium has been removed, solidified, and is stored on-site. The

facility has been placed in a long-term, low-cost surveillance and maintenance

condition.

The 618-10 (300 North) Waste Burial Ground is approximately 3.5 miles south of CGS. DOE has initiated remediation activities at the site.

The 618-11 (Wye) Waste Burial Ground is directly west of CGS, outside of Energy Northwest leased land, but within its 1950-meter exclusion area radius and security

perimeter. The site received low- to high-activity waste, fission products, some

plutonium-contaminated waste, and non-radiological hazardous waste from March 1962

to December 1967 from the Hanford 300 Area. The waste is buried in 3 trenches, 50 Vertical Pipe Units (VPUs), and four caissons. The site was covered with an overburden of soil when it was closed. The surface was stabilized in 1982 with an additional 2 ft of soil. Since surface stabilization, activities at the site have been limited

to monitoring and surveillance. DOE completed non-intrusive surveillance and characterization activities at the site in 2011 to obtain data information and information

for planning remediation activities.

The DOE 300, 200 East, and 200 West Areas are located within a 10-mile radius of the site.

The current waste management activities (storage, disposal, and treatment) conducted in these

areas are discussed in Table 2.2-1. The 300 Area is approximately 7 miles southeast of CGS.

The only hazard presented to CGS from this site is from the spent nuclear fuel and other

radioactive material stored there. There is an unknown quantity of miscellaneous reactor fuel

material in the 300 Area. This quantity is not publicly available information.

The DOE 200 East and 200 West Areas are approximately 10 miles northwest of CGS.

Originally these facilities were constructed to support the extraction of weapons grade

plutonium for the defense program. However, as the Hanford mission has changed from production to environmental cleanup, so has the purpose of the facilities discussed. This change in mission has, in some cases, resulted in a change in the hazards presented to CGS

plant site and personnel.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 2.2-6 A private (non-DOE) low-level radioactive waste disposal area is adjacent to 200 East Area. There are also plans to build private waste vitrification facilities adjacent to 200 East. These

facilities are also discussed in Table 2.2-1.

Several plutonium production reactor facilities are located approximately 20 miles north-northwest of CGS. All of the reactors were water cooled, graphite moderated. The last operating reactor, the N Reactor, was permanently shut down in 1991. The N Reactor also

provided steam for the Energy Northwest Hanford Generating Station until 1987. The fuel has been removed from the reactors for storage or treatment. The current activities at these reactor

sites are also discussed in Table 2.2-1.

The nearest petroleum product storage tanks are located 22 miles southeast of the site:

approximately 23 million gal capacity at the Chevron Pipeline Company and approximately

20 million gal capacity at the Tidewater Barge Lines.

2.2.2.2 Description of Products and Materials

The existing Hanford Site railroad track (owned by the DOE and operated by a private

contractor in support of the Hanford Operations), and the CGS, the WNP-1, WNP-4, and the

FFTF railroad spurs all run within the exclusion area of the plant site. Shipments of large quantities of hazardous materials on this track that existed during initial licensing of CGS are

no longer made (Reference 2.2-4).

The DOE has no plans for railroad shipments of explosives in the foreseeable future. However the DOEs Richland Operations Office has agreed to notify Energy Northwest prior to transporting any explosive shipments of more than 1800 lb past CGS (Reference 2.2-5). Energy Northwest will provide an analysis to the NRC of the potential consequences prior to

the start of such shipping (Reference 2.2-6).

Hazardous material is also transported on Hanford Route 4 by DOE. Chlorine is the only

material of concern transported on Route 4 with the potential for impacting CGS operation.

Section 6.4.4.2.1 provides additional information on control room habitability assessments for CGS.

The Yakima Training Center, a sub-installation of Fort Lewis, is 30 miles northwest of the site. The center consists of 327,000 acres. The center provides training facilities and logistical

support and it is used for firing of all types of ordnance, both in a direct mode and by indirect

artillery and mortars. Weapons to 155mm are fired. This type firing occurs frequently. Other types of live ordnance use at the center include aerial delivery by high performance aircraft of ordnance to include 2,000-lb bombs, helicopter weapons which include automatic weapons and

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 2.2-7 2.75-in. folding fin rockets, and anti-aircraft mi ssiles. These latter activities are significant as to occurrence. The majority of the ordnance impacts a 20,000-acre area which is generally located in the central portion of the center. All activities are confined to the geographical limits of the center and/or its restricted air space unless special arrangements are made with affected agencies. Mechanized units (i.e., tanks and armored personnel carriers) from Fort Lewis and reserve components conduct extensiv e maneuvers on all accessible areas of the Training Center and use specially designed ranges to practice fi ring their weapons. Infantry and engineer units that support the mechanized un its also train at the center. Training activity is greatest from March to November. War games sometimes involve troop and equipment deployment at the Richland Airport and al ong Highway 243 west of Vernita Bridge.

Helicopters may fly near the Hanford Site, or military vehicles may travel over Highway 240 (Reference 2.2-7).

2.2.2.3 Pipelines

There are no commercial oil or gas pipelines in th e vicinity of CGS. The nearest major natural gas transmission pipeline to the site is about 12 miles. A 20-in. gas transmission line of the Northwest Pipeline Corporation is located east and essentially paralle l to U.S. Highway 395 between Pasco and Ritzville, Washington. A second pipeline system consisting of parallel

36-in. and 42-in. lines, owned by Pacific Gas Transmission Company, passes through Wallula, approximately 24 miles from the site (Reference 2.2-8). These distances eliminate any potential hazard to plant operations due to a na tural gas fire or explosion. The Energy Northwest Hydrogen Storage and Supply Facility is located 0.6 miles south-southeast of the plant and is connected to the plant with a 2-in NPS gas pipeline. The pipeline runs north from the facility approximately 400 ft east of the plan t and then turns and r uns west approximately 400 feet north of the plant then south approximate ly 200 ft west of the plant to its connection point on the west side of the Turbine Building.

Fire and explosion risk s to the plant involving this pipeline are discussed in Appendix F and Section 3.5.1.5.

2.2.2.4 Waterways

Makeup water inlet structures are located in the Columbia River 315 ft from the shoreline at low river flow (36,000 cfs; el.

341.73 ft) at river mile 351.75.

A significant amount of Columbia River barge tr affic moves as far upstream as the Ports of Pasco and Kennewick. Also, a docking facility established by the Port of Benton in North Richland (approximately 9 miles downstream of the CGS site) is accessible by barges with a maximum 16 ft of draft (normally 2500 to 3000 tons

). The first use of this facility was in April 1973 when the FFTF reactor vessel was off-l oaded. Traffic to the North Richland dock is very infrequent in comparison to that in Pasco and Kennewick due to the lack of large industrial concerns in the region between Richla nd and Priest Rapids Da

m. This facility is most often used to off-load dismantled nuclear co mponents. On several occasions in the past, C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 2.2-8 lightly loaded barges have transported material to the vicinity of the Hanford Site. This required maintenance of an ade quate flow from Priest Rapids Dam during the transit period.

2.2.2.5 Airports

Three commercial airports are w ithin 20 miles of CGS. The closest is Richland Airport 11 miles south of the plant. This general avia tion airport has two 4000-ft runways, one with a 010/190 orientation and the other with a 070/250 orientation. Visual flight rule landings are standard for Federal Aviation Administration (FAA) non-control-tower airports.

The Tri-Cities Airport 17 miles southeast near Pa sco is the largest airport within 40 miles. The FAA operates the air traffic tower and airport radar approval control facility. The airport has two 7700-ft crossing runways with 120/300 and 030/210 orientations. The latter has a 4430-ft parallel runway. Runway 30 has a very high frequency omnira nge (VOR) instrument approach and Runway 21R has an instrument la nding system and is an instrument approach runway.

The Vista Airport operated by the Port of Kenne wick is a general aviation airport located 18 miles south-southeast. It has a 4000-ft runway with a 20/200 orientation. All operations are under visual flight rules.

Information relative to the flight paths and ac tivity at these three co mmercial airports, the Yakima Training Center, and the nearby priv ate airstrips is discussed in Section 3.5.1.6.

2.2.2.6 Projection of Industrial Growth

There is no projected growth of waterway tra ffic nor plans for oil and gas pipelines within 10 miles of CGS.

2.2.3 EVALUATION

OF POTENTIAL ACCIDENTS

2.2.3.1 Determination of Design Basis Events

Energy Northwest has investigated the resistance of plant structures to explosions. The reactor building is a reinforced-concrete structure up to the refueling floor and is designed to withstand the worst probable combination of wind velocity and associated pressure drop due to a design basis tornado. A differential pressure of 3 psi between the exteri or and interior of the building is also considered in the design. At its near est point, the railroad is 510 ft from the reactor building.

From the above criteria, it has been determined th at the reactor building can resist an explosion of 20,000 lb of dynamite on a railway car 510 ft from the reactor building. The performance of the reactor building structure for this blast loading condition will be similar to that for the

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 2.2-9 original design basis tornado loading condition ba sed on the 3 psi differential pressure used in the original tornado analysis.

In the unlikely event of an explosion or fi re on the railroad affec ting the 115-kV shutdown power supply, the 230-kV power supply or the di esel generators would fulfill that function.

It is extremely unlikely that an explosion or fire on the mainline railroad would compromise

the safe shutdown of the facility. As noted in Section 2.2.2.2 , DOE has no plans to ship explosives on the railroad, and the agency will notify Energy Northwest prior to the shipment of explosives in a quantity greater than 1800 lb. Energy Northwest Security controls access to the rails that pass near the plant. The only explosives on the Hanford Site are small arms munitions. As described in Section 2.2.2.2 , this represents no hazard to the operation of CGS.

The Yakima Training Center does not endanger the site. Hydrogen gas stored in the gas bottle storage building and in a trailer parked adjacent to the gas bottle storage building will not pose

any fire or explosion problem because of the light weight properties and dispersal qualities of the gas and the distances (approximately 400 ft) between the storage areas and any safety-related equipment. Hydrogen gas stor ed at the Hydrogen Storage and Supply Facility (HSSF) and transported to the plant by pipeline does not pose a significant fire or explosion risk to the plant as discussed in Appendix F and Section 3.5.1.5. Habitability considerations for the control room and quantities of hydrogen ga s stored in/adjacent to the gas bottle storage building and at the HSSF are discussed in Section 6.4.4.2.3.

Table 2.2-1 summarizes the potential events at the Hanford Site facilities that could present a radiological or chemical hazard or hazardous situa tion to the continued sa fe operation of CGS.

The cesium and strontium capsule s stored at the Waste Encap sulation and Storage Facility (WESF), the fuel stored at the K Basins, and the high level waste stored in the tank farms present contributions to risk at CGS due to presence of 137 Cs, 90 Sr, and 241 Am. However, based on consideration of the radionuclid e inventory at risk, the ability to transport this inventory, and the proximity of the storage facility, the ri sk is dominated by the inventory stored at WESF. The probability of the loss of cooling in the capsule storage pool is extremely low, but the potential dose to unprotected CGS personnel due to the release is significant. Any required evacuations would be performe d as discussed in the Emergency Plan. The design basis

accident at WESF would not result in a conditi on at CGS which would challenge the criteria established in 10 CFR 100.

In each event evaluated, the radiological dose resulting from particulate releases would be adequately filtered by the control room HEPA filters, mitigating any challenge to the habitability of the control room. None of th e facilities present a chemical exposure risk to CGS. Radiological exposures fo r postulated events on the Hanfor d Site are characterized by the contribution from gaseous radionuclides because of the short half-life of 131 I and because the other noble gases were released during the spent fuel processing.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-038 2.2-10 The Yakima Training Center, discussed in Section 2.2.2.2 , is used for military maneuvers and weapons training and is the only significant military activity in the vicinity of the Hanford Site.

The only weapon currently in use at the Yakima Training Center known to present a hazard to the Hanford Site is the multiple launch rocket system (MLRS). With a range in excess of 25 miles, the MLRS could potentially impact the CGS site. However, the MLRS is only fired

from the perimeter of the Yakima Training Center into a centrally located impact zone and is only fired with dummy warheads. Given this in formation, additional safety features, and the administrative controls in place at the Yakima Training Center, a weapons accident having an impact on CGS is very improbable.

As stated in Section 2.2.2.1 , confinement of all radioactive materials at the ISFSI is provided by the required use of NRC certified spent fu el storage casks listed in 10 CFR 72.214.

Pursuant to the 10 CFR 72.212 report, evaluati ons performed in support of the ISFSI have demonstrated the reactor site-specific parame ters are bounded by the safety analysis for the generically approved cask. Acco rdingly, activities associated w ith the facility do not adversely impact operation of CGS (Reference 2.2-10).

Brush fires have occurred on the Hanford Site and have presented no potential hazard to

existing facilities. Areas adjacent to CGS major buildings and auxiliary facilities are maintained to prevent weed growth by landscaping, gravel, ground cover, and weed control spraying. The Hydrogen Storage and Supply Facility (HSSF) is landscaped with gravel beyond the perimeter of the site, exceeding the c ode required clearance distance, to keep the area free from dry vegetation and combustible mate rials. These or similar methods of weed control minimize brush fire hazards to CGS facilities.

The potential effects of fires that involve mate rials used in the operation of the plant are discussed in Appendix F.

The formation of unconfined vapor clouds cause d by the accidental release of flammable or toxic liquids or vapors stored at the plant site is discussed in Section 6.4 and addressed by the Emergency Plan.

The non-safety-related makeup wate r intake consists of two sets of paired perforated pipe sections. One set is capable of supplying the full makeup water requireme nts of the plant. Extreme low river flow (36,000 cfs) will provide about 0.5 ft of water over the top of the

intake pipes. The probability of damage to both sets of intakes as a result of a pleasure boat or barge accident is extremely remote given the in frequency of both extreme low flows and large boat and barge traffic. In the unlikely event that such an accide nt might occur, destruction of the makeup water intake structure would be comparab le in effect to loss of offsite power to the makeup water pumps. The Seismic Category I sp ray ponds provide for 30-day cooling without makeup. This is ample time to restore makeup from either the river or wells.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 2.2-11 There are no upstream industrial facilities for which waterborne deliveries of significant quantities of petroleum products, corrosive ch emicals, or other hazardous materials are expected. Fuel oil, diesel oil, acids, and caus tics are stored at the N-Reactor site. The oil storage facilities are protected by dikes, and th e chemical storage facilities are far enough from the river to avoid direct disc harge. Thus, there is no possi ble hazard to the plant due to spillage of such materials into the river.

There are no upstream releases which may be corrosive, cryogenic, or coagulant.

2.2.3.2 Effects of Design Basis Events

As discussed in Section 2.2.3.1 , the activities of nearby industria l, transportation, and military facilities will have no adverse effect on the plant.

2.

2.4 REFERENCES

2.2-1 Holmes, D. B., Energy Northwest, personal communication with Steve Burnam, Site Infrastructure Division, Department of Energy, July 31, 1998.

2.2-2 Presidential Proclama tion 7319, Establishment of the Hanford Reach National Monument, 65 FR 37253, June 9, 2000.

2.2-3 Chasse, J. P., Energy Northwest, personal communication with B. J. Rokkan, Safeguards and Security Division, DOE, Richland Operations Office, December 6, 1977.

2.2-4 Note, D. A. Marsh, Westinghouse Hanford, to D.

E. Larson, Supply System, dated January 22, 1993.

2.2-5 Memorandum, C. A. Hansen, Assi stant Manager for Waste Management, DOE-RL to Alice Q. Murphy et al

., 98-WPD-032, February 20, 1998.

2.2-6 NUREG 0892, Safety Evaluation Report Related to the Operation of WPPSS Nuclear Project No. 2, S ection 2.2.1, March 1982.

2.2-7 Arbuckle, J. D., Energy Northwest, personal communication with J. Reddick, Executive Officer, Yakima Training Center, July 29, 2003.

2.2-8 Hosler, A. G., Contact with Lo cal Organizations to Support CSB SAR Chapter 1 (Memo 042AGH.96 to Canist er Storage Building Report File), Science Applications Internati onal Corporation, May 6, 1998.

C OLUMBIA G ENERATING S TATION Amendment 64 F INAL S AFETY A NALYSIS R EPORT December 2017 LDCN-16-038 2.2-12 2.2-9 Deleted.

2.2-10 Energy Northwest, I ndependent Spent Fuel Stor age Installation 10 CFR 72.212 Evaluation, Docket Number 72-35, Revision 1, September 2002.

Table 2.2-1 Hanford Site Nuclear Facilities Facility Description Hazard Design Basis Event Impact on CGS C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-11-027 2.2-13Fast Flux Test Facility (FFTF) Ref: Surveillance and

Maintenance Plan for FFTF, Rev. 0, DOE/RL-2009-26 Deactivated sodium cooled breeder reactor Activated solid sodium

1) 180,000 lb sodium spill hr dose at 1.5 mi = 0.015 mrem, 24-hr dose at 4.5 mi = 0.26 mrem Particulate release, effectively mitigated by distance DOE 618-10 (300 North) Waste Burial Ground

Ref: Final Hazard Categorization for the 618-10 Burial Ground Remediation, Rev.2, February 2011, WCH-390 Disposal site with broad spectrum of low- to

high-level solid radioactive wastes buried in caissons or

pipe Radioactive waste Caisson penetration with fire - unmitigated dose at 5 km = 10.2 mrem Particulate release, effectively mitigated by distance (>5 km)

DOE 618-11 (Wye) Waste Burial Ground Ref: 618-10 and 618-11 Waste Burial Grounds Basis

for Interim Operation, Rev.

1, WCH-183 Disposal site with broad spectrum of low- to high-level solid radioactive wastes and non-radiological hazardous materials buried in caissons or pipe Radioactive waste inventories, primarily Cs-137, Sr-90, and Pu-239

and non-radiological hazardous materials bounded by beryllium Caisson penetration with fire and explosion: control room doses are less than 0.1 rem; beryllium oxide

concentration of 4.6 x 10-3 mg/m 3 at 100 m from 618-11 site boundary Particulate release, effectively mitigated by credited 618-11

Waste Burial Ground project controls. The soil overburden covering the VPUs and caissons in the 618-11 Waste Burial Ground is credited for

reducing releases and is designated as a passive design feature. No missiles are postulated for this event.

Table 2.2-1 Hanford Site Nuclear Facilities (Continued)

Facility Description Hazard Design Basis Event Impact on CGS C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011 2.2-14 B Plant Ref: B Plant Basis for Interim Operation, March 6, 1997, HNF-SD-BIO-003 Process to remove cesium and strontium from radioactive waste, deactivated, currently in surveillance and maintenance mode Residual radionuclide inventories on cell filters (137 Cs, 90Sr, and 241 Am) Flooding cell 291-B HEPA filters -

0.368 rem max. public dose Particulate release effectively mitigated by distance Plutonium-Uranium Extraction Facility (PUREX). PUREX End State Basis for Interim Operation (BIO) 1997, HNF-SD-CP-15B-004 (Draft) Currently shut down, in preparation for decommissioning and decontamination Residual plutonium and uranium contamination Design basis earthquake, dose @ 100 m - 1.9 rem; 12 km -

7.4 x 10-4 rem Particulate release, mitigated by distance Plutonium Finishing Plant (PFP) Ref: Plutonium Finishing

Plant Final Safety Analysis

Report, 1995, WHC-SD-CP-SAR-021 Receipt and storage of SNM, reactive material stabilization, radioactive and mixed waste handling Stored SNM, and residual plutonium contamination Design basis earthquake, 8-hr dose of 15.2 rem @550 m 24-hr dose of 0.31 rem @12,500 m Particulate release, mitigated by distance Tank Waste Remediation System (TWRS) Facilities Ref: Tank Waste Remediatio n System Basis for Interim Operation, 1997, HNF-SD-WM-BIO-00, Revision 0 Mixed radioactive and chemical wastes storage in 149 single shell tanks (SST)

and 28 double shell tanks (DST) in 12 tank farms

Associated support facility:

242-A-Evaporator SSTs contain combinations of sludge, saltcake, and interstitial and pooled liquids

DSTs contain liquid and slurry waste with small amounts of sludge

Table 2.2-1 Hanford Site Nuclear Facilities (Continued)

Facility Description Hazard Design Basis Event Impact on CGS C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011 2.2-15 Tank Waste Remediation System- Project (TWRS-P) (private facilities, proposed for construction in 2000) Ref: DOE/RL-96-0006 Vitrify low level, high level, and transuranic mixed waste High level radioactive waste Requirements for authorization limit to less than 25 rem/event in an accident. Releases are projected to be airborne particulate. HEPA filter system would protect CGS control room operators, in accordance with the limits of GDC 19 Waste Encapsulation and Storage Facility (WESF)

Ref: Waste Encapsulation and Storage Facility Basis for Interim Operation, 1997, HNF-SD-WM-BIO-002 Conversion process for cesium and strontium has halted Cesium chloride and strontium fluoride salts, encapsulated in double-walled metal containers stored in water-filled cooling basin Capsule rupture following loss of water from storage pool - 24 hr exposure to public, 9 rem Control room exposure mitigated by

HEPA filters; potential evacuation of other personnel Low-Level Waste Disposal Site (Private) Buried storage of low-level radioactive waste in lined containers Low-level buried waste, monitored as required by NRC license No credible event None Canister Storage Building (CSB)

Ref: Letter; DOE to

H. J. Hatch, Flour Daniel Hanford, 28 May 97 Storage of spent nuclear fuel (SNF) from the K Basins in sealed multi-canister overpacks (MCO) 2100 metric tons of spent fuel, from production reactors Requirements for authorization limit to less than 5 rem/event in an accident. Releases are projected to be airborne particulate HEPA filter system would protect CGS

control room operators, in accordance with the

limits of GDC 19 Cold Vacuum Drying Facility (CVDS) Ref: Letter; DOE to

H. J. Hatch, Flour Daniel Hanford, 28 May 97 Draining and vacuum drying to remove water from MCOs in preparation for interim storage at CSB Spray release Requirements for authorization limit to less than 5 rem/event in an accident. Releases are projected to be airborne particulate. HEPA filter system would protect CGS control room operators, in accordance with the limits of GDC 19

Table 2.2-1 Hanford Site Nuclear Facilities (Continued)

Facility Description Hazard Design Basis Event Impact on CGS C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORTDecember 2011 2.2-16B, C, D, DR, F, and H Reactors (shutdown since 1969)

Ref: WHC-EP-0619, Vol. 1. Single-pass, water-cooled, graphite-moderated reactors The onsite hazards involve personnel injury, primarily falls and electric shock. No significant offsite risks No credible event None K East and West reactors (shutdown since 1971)

Ref: WHC-EP-0619, Vol. 1. Single-pass, water-cooled, graphite-moderated reactors The onsite hazards involve personnel injury, primarily falls and electric shock. No significant offsite risks No credible event None K East and West Basins Ref: WHC-EP-0619, Vol. 1. Storage basins for spent fuel, some severely degraded Approximately 2100 metric ton inventory of irradiated reactor fuel Dropping and overturning of a transfer cask containing reactor fuel. (Bounding event; but transfer is administratively prohibited)

None N reactor (shutdown since 1987) and N Basin

Ref: WHC-EP-0619, Vol. 1. and BHI-00866, Rev 0 A pressure tube, water-cooled, graphite-moderated reactor

with fuel assemblies

removed for decontamination and decommissioning The onsite hazards involve personnel injury, primarily falls and electric shock. No significant offsite risks No credible event None

NORTHSUNNYSIDE PROSSER Benton County RICHLANDPASCO COLUMBIA COLUMBIA RIVER AG CARR KENNEWICKVISTA FFTF COLUMBIA AG 2 BASIN CITYWye BarricadeMcWHORTER Hanford Site BoundaryMATT AW A Grant County Christensen Bros.DESERT AIR 301 V187°Adams CountyTAGGARES FIELD KENTFARMS OTHELLO CONNEL CITY Franklin County GREEN ACRES 332°YakimaTraining Center CGS 200West 200 East SLINKARD 269°24 243 24 17 240 4S VR1350 V281 IR329 IR329 IR329 VR1351 VR1360 VR497 V204 12 395 395 395 82 182 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-1 2.3 METEOROLOGY The italicized information, including associat ed tables and figures, is historical and was provided to support the application for an operating license.

2.3.1 REGIONAL

CLIMATOLOGY

• 2.3.1.1 General Climate The site is located in a mid-latitude semi-arid (steppe) climatic region in the Lower Columbia Basin which is the lowest elevation of any part of central Washington. A major factor influencing this climatological re gion is its location in the con tinent, well away from the windward coast and protected to the west by the 4,000 to 7, 000-ft average elevation Cascade Mountains. Dominant air masses affecting the region are of maritime polar origin as modified by the presence of these mountains. Modified continental tropical and polar air masses also periodically affect the climate. In winter, there is a succession of cyclones as the westerlies and the polar front prevail in these latitudes. The mountain barriers commonly induce these storms to occlude by delaying air mass movemen

t. Fewer frontal passages occur during the summer months since subtropica l oceanic high cells reach th eir highest latitudes thereby diverting cyclonic storms poleward. Along the eastern margin of the Pacific anticyclone, an

out-flow of stable subsiding air br ings distinctly drier conditions to the North American Pacific coast.

The regional temperatures, precipitation, and winds are greatly affe cted by the presence of the mountain barriers. The Rocky Mountains and ranges in Southern British Columbia are effective in protecting the inland basin from the more severe wint er storms and associated cold polar air masses moving southward across Canada.

Occasionally, an outbreak of cold air will pass through the Basin and result in low temperatures or a damaging spring or fall frost.

Maritime polar air traveling eastward from the coastal zone cools as it rises along the western slope of the Cascade Range.

These orographic effects cause heavy precipitation on the windward and light precip itation on the leeward slopes. Th e prevailing westerly. winds are normally strongest during winter and spring due to the presence of cyclonic scale disturbances and associated frontal activity.

During those months, foehn or chinook winds (a warm dry

• This section is based on record s kept at the Hanford Meteorol ogy Station (14 miles northwest of the site, elevation 733 ft MSL) from 1945 to 1980 (2) and 100-N area sites (1) (supplemented with precipitation and temperature data taken by U.S. Weather Bure au cooperative observers at a site about 25 miles north of the present station location during the period from 1912 to 1944 (2,3) and regional c limatological data gathered during th e period from 1931 to 1960). (4)

Other references are as indicated.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-2 wind on the lee side of a mountain range; the warm th and dryness of the ai r is due to adiabatic compression upon descending the mountain slope s) occur whenever cycl onic circulation is sufficiently strong and deep to force air completely across th e cascades in a short period of time. At other times during the winter, warm front occlusions can fo rce moist air over the Cascade Range. The mixing of this moist air with relatively cooler air in the Basin results in considerable cloudiness and fog. The percent of possible sunshine ranges from 20 to 30 percent in winter, 50 to 60 percent in spri ng and fall, and 80 to 85 pe rcent in mid-summer.

Because the site is in the rain shadow of these mountains , annual average precipitation decreases from about 100 inches near the summit of the Cascades to about 6 or 7 inches in the Basin. Approximately 70 percent of the annu al total precipitation occurs from November through April and about 10 percent occurs during July through September. Rainfall amounts are normally light in the summer and gradually in crease in late fall, reaching a peak of about one inch each month in midwin ter due to cyclonic storm and front al activity. Rainfall amounts decrease in Spring, increase somewhat in June, and again sharply decrease in July. During mid-summer, it is not uncommon to have 3- to 6-week periods w ith trace rainfa ll. There are only two occurrences per year of 24-hour amounts of 0.50 inch or more, while occurrences of 24-hour amounts of 1.00 inch or more number onl y four in the entire 25 years of record (1946 to 1970). One of these was the record storm of October 1 through 2, 1957, in which rainfall totaled 1.08 inches in three hours, 1.68 inches in six hours, and 1.88 inches in twelve hours.

At the other extreme, there have been 81 consecutive day s without measurable rain (June 22 through September 10, 1967), 139 days with only 0.18 inch (June 22 through November 7, 1967), and 172 days with only 0.32 in ch (February 24 through August 13, 1968).

About 45 percent of all precipitation during the months of December, J anuary, and February is in the form of snow. Regional annual tota l snowfall amounts have ranged from less than 1/2 inch in 1957 to 1958 to 56.1 inches for the winter of 1992-1993; the annual average total is about 14 inches.

Snow rarely remains on the ground longer than two to four weeks or re aches a depth at any time in excess of four to six in ches, as rapid melting, which ofte n contributes to local stream flooding, can occur from rain or Chinook winds. The record greatest depth of 24.5 inches occurred in February 1916.

Thunderstorms have been observed in the area in every month except November. Although severe ones are rare, lightning strikes have occasionally ign ited grass fires which burned thousands of acres of the Hanford Reservation and resulted subsequently in considerable wind erosion of soil. The most notable of these oc currences were in August 1961, July 1963, and July 1970, and August 1984.

The continental-type climate not only affects prec ipitation in the Basin but also results in wide ranges and variations in annual temperature conditions. While the regional annual average temperature is about 53°F, the coldest month, January, has a mean of about 29°F; the C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-3 warmest month, July, has a mean of about 76

°F. Although the pres ence of the cascades contributes to the wide differences in monthly average temperatures, other mountain ranges shield the area from many of the arctic surges, and half of all winters are fr ee of temperatures as low as 0°F. However, six winters in 58 of record have contributed a total of 16 days with temperatures of -20 F or below; and in January to Febr uary 1950, there were four consecutive such days. There are ten days of record when even the maximum temperature failed to rise above zero. At the other extreme , in the winter of 1925 to 1926, the lowest temperature all season was +22 F. Although winter minima have varied from -27 F to +22 F, summer maxima have varied only from 100°F to 115°F. However, there is considerable variati on in the frequency of such maxima. In 1954, for example, there was only one day with a maximum as high as 100°F. On the other hand, there have been two summers (1938 and 1 967) when the temperature went to 100°F or above for 11 consecutive days.

Although temperatures reach 90°F or above on about 56 days a year, there are only about seven annual occurrences of over night minima 70°F or above.

The usual cool nights are a result of gravity winds.

The channeling of air by the Cascade Mountains and surrounding terrain produces a prevailing WNW and NW regional flow. Local topographic features can cause ot her channeling effects and formation of local diurnal wind circulation systems which produce a greater degree of variability in winds at locations within the Basin. For example, the Columbia Generating Station (CGS) site experiences a bimodal wind direction distribution from approximately south and also northwest; at the H anford Meteorological Station (HMS) about 14 miles northwest, the direction distribution displays a single pe ak at approximately WNW to NW (refer to 2.3.2).

Drainage (gravity) winds channeled by topographic features produce a marked effect on

diurnal range of wind speed and cause the highest monthly average speeds of about 9 mph to occur during the summer months. In July, for example, hourly average speeds range from a low of 5.2 mph from 9 to 10 a.m.

to a high of 13.0 mph from 9 to 10 p.m. In contrast, the corresponding speeds in January are 5.5 and 6.3 mph. These warm season diurnal winds, resulting from relativel y cold air draining from the Cascade Mountains, occur in response to pressure gradients created betw een surface-heated warm, dry bas in air and cooler air situated over the mountains and coastal region. This favors an outbrea k of stronger winds during the afternoon and evening hours. Althoug h the gravity wind occurs with regularity in summer, it is never strong unless reinforced by frontal activity.

In June, the month of highest average speed, there are fewer instances of hourly averages exceeding 31 mph than in December, the month of lowest average speed. A complete summary of the monthly averages and extremes of climatic elements at the Hanfor d Reservation appears in Table 2.3-1.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-4 2.3.1.2 Regional Meteorological Conditions for Design and Operating Bases 2.3.1.2.1 Severe Weather Phenomena

2.3.1.2.1.1 Heavy Rain, Snow, and Ice. Glaze is a coating of ice, generally clear and smooth, but with some air pockets. It is formed on exposed object s by the freezing of super-cooled drizzle or rain drops. Glaze is denser, harder, and more transparent than either rime or hoar frost. Although th e record shows an average of seven glaze days per year, many of these cause little or no inc onvenience to the public. Two out standing exceptions occurred on February 11 to 12, 1954, and on November 23 to 24, 1970. Ther e was serious disruption to Hanford traffic in each instance although there was no known damage to transmission lines. In each instance, rising temper atures soon melted the ice.

Precipitation frequency (rain and snow), intensity, and quantity st atistics are presented in Figures 2.3-1 and 2.3-2 and Tables 2.3-2 and 2.3-3. For the winter of 1992-93, the following snowfall records were set: grea test winter snowfall (56.1 inches

); most days with greater than 1, 6, and 12 inches on the ground (71, 41, and 9, respectively); and greatest 24-hour snowfall (10.2 inches) on February 18 and 19. Probable maximum pr ecipitation is given in 2.4.3.1.

2.3.1.2.1.2 Thu nderstorms and Hail. Thunderstorms may occu r during any month of the year at Hanford. A thunderstor m day is one in which thunder is heard. If a thunderstorm should begin in late evening and last past mi dnight, it is counted as two thunderstorm days even though only one storm event occurred. Simila rly, should there be two or more distinct thunderstorms in a day - and this sometimes happens - it is c ounted as a singl e thunderstorm day. The table below shows the mo nthly frequency of thunderstorms.

HMS TH U N DERSTO R M DAYS: 1945-1970

J F M A M J J A S O N D SUM Total 0 1 7 18 53 64 46 54 24 5 0 0 272 Average 0 # # 1 2 3 2 2 1 # 0 0 11

% of Total 0 # 3 7 19 23 17 20 9 2 0 0 100

1. = Less than 0.5

Although the table above shows 0 for the months of November through January, a thunderstorm occurred at HMS on December 22, 1971. In Richland, one occurred on January 18, 1953. However, th e thunderstorm season essentially includes only the months of April through September. Alt hough the average is eleven days pe r year, the number has varied from three to twenty-three.

In June 1948, there were ei ght thunderstorm days during the month; and this record was re peated in August 1953. The r ecords show that cold fronts probably constitute the greatest single cause of thunderstorms at HMS. During the years of 1947 to 1955, 43 percent of all thunderstorm days during the months of May through August C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-5 were directly associated with cold frontal passages. On several occasions (notably on August 7, 1953), lightning has struck the HMS tower.

Peak gust data are not available for the 50-, 200-, and 400- ft levels prior to 1952. Of the

185 thunderstorm days occ u rring during the period of 1952 to 19 7 0, the speed classification of

peak gusts on these days is as follows:

Number of Cases  % of Total mph 50 ft 200 ft 400 ft 50 ft 200 ft 400 ft < 21 18 9 5 10 5 3 21 -30 75 45 42 40 24 23 31 -40 63 80 73 34 43 39 41 -50 23 34 46 12 19 25 51 -60 4 11 12 2 6 8 61 -70 1 4 5 1 2 3

> 70 1 2 2 1 1 1 185 185 185 100 100 100

Precipitation was not m e asured during 1945 and 1946 in wh i c h 26 thunderstorm days

occurred. During the period of 1947 to 197 0 , 246 thunderstorm days we re recorded. The

daily precipitation distribution d u ring these days was as follows:

Amount - Inches Number of Cases

% of Total None or trace 110 45 0.01 - 0.10 87 35 0.11 - 0.25 29 12 0.26 - 0.50 15 6 > 0.50 5 2 246 100 Precipitation intensities are defined in Reference 2.3-5.

The record for rainfall intensity during a thunderstorm is 0.55 inch in 20 minutes (1.65 inches per hour) on June 12, 1969. This storm included hailstone s of 1/4-inch diameter.

Hail was reported on fourteen, or 5 percent, of the total thunde rstorm days. Blowing dust or dust was reported on sixteen thunderstorm days and both hail and blowing dust or dust on six days.

Hail is a rare phenomenon at Hanford. For a ll years of record, hail ha s not occurred more than twice in any year. Of the 272 thunderstorm days from 1945 to 1970, hail was reported on fourteen or 5 percent of these days. Hail was also reported on two days without occurrence C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-6 of either a cold frontal passage or a thunderstorm on the same day. The distribution by months of days on which hail occurred is as follows:

J F M A M J J A S O N D Total Number 0 1 1 4 2 1 2 2 1 0 0 0 14 % of Total 0 7 7 30 14 7 14 14 7 0 0 0 100 Where size was reported, all except two reports i ndicated sizes in the 0.2-to 0.3-inch range.

The exceptions were May 26, 1954, and July 1, 1955, when the size reported was 0.4 inch.

There is no known case of local damage from hail.

2.3.1.2.1.3 Tornadoes

The State of Washington experie nces, on the average, less than one tornado each year. Within a one hundred mile radius of the site, only f ourteen tornadoes have b een reported since 1916.

These tornadoes are listed in Table 2.3-4. Of these fourteen reco rded tornadoes, only five had any damage associated with them. A more extensive survey of tornadoes in the three northwestern states (Washington, Oregon, and Idaho) was performe d by Fujita (6). His results indicate that tornadoes and hailstorms in this area occur primarily in "a lleys". The locations of these "alleys" are shown in Figure 2.3-3 along with locations of tornadoes which have been recorded in the tri-state area during th e twenty-year period from 1950 to 1969.

Jaech (7) has analyzed the data of Fujita (6) to determine the probability of a tornado striking the Jersey Nuclear Company Fuel Facility (now Siemens Power Co rporation), which is located about eight miles from the si te. His analysis estimates the probability of occurrence of a tornado in the vicinity of the Exxon site as six chances in a million during any given year or about one chance in four thousand during a forty-year plant life.

The peak tornado wind velocity estimated for the site is 214 mph (Reference 2.3-7). This includes an estimated maximum rotational and translational wind velo cities (at a 95-percent confidence level). Daubek (Reference 2.3-8) estimates the maximum tr anslational velocity to be 30 mph. The maximum pressure drop in the center of the tornado relative to the environment is estimated to be up to 1.5 psi (Reference 2.3-6). These values are equal to or le ss than those tornado parameters listed for a Class III region (which includes the site location) in Regulatory Guide 1.76, issued April 1974. Prior to the issuance of this guide, CGS was designed to withstand some of th e most stringent NRC tornado criteria presented for a site located within a Class I region.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-7 A comparison between criter i a used for CGS and t hose applicab l e to C l a ss I and III regions are given belo w:

Design Bas i s Tornado Character i s t i cs

Maximum Wind Rotational Translational Speed (mph)

Speed (mph) Speed (mph)

Maximum Minimum Class I Region 360 290 70 5 Class III Region 240 190 50 5 CGS 360 300 60

- Radius of Maximum Rotational Speed (ft)

Pressure Drop (psi) Rate of Pressure Drop (psi/sec)

Class I Region 150 3.0 2.0 Class III Region 150 1.5 0.6 CGS 264 - 880 3.0 1.0 Wind and tornado loading criteria used in th e CGS structural design are discussed in Section 3.3.

2.3.1.2.1.4 Strong Winds. The Hanford region experiences high wind speeds due to squall lines, frontal passages, strong pressure gradi ents and thunderstorms. The Hanford Reservation has experienced only one recorded tornado (June 1948) and has not been known to be affected by typhoons. No comp lete statistics are readily available whic h present frequency of occurrence of high winds produced or accompanied by a particular meteorological event.

However, the highest winds produced by any cause are tabulated for HMS in Tables 2.3-5 and 2.3-6. Figure 2.3-4 indicates the return proba b ility of any peak wind gust, again due to any

cause.

The speed-direction summary (Table 2.3-6) shows that daily peak gus ts of at least 40 mph have occurred from all but four of the sixteen co mpass points indicated. The SW octant, however, accounts for 65 percent of such cases. The SSW octant accounts for 83 percent of daily peak gusts of 50 mph or over and 100 percent of those 60 mph or over. Since WNW and NW are the most frequently observed direc tions at HMS, they account for almost half of all daily peak gusts. However, less than 3 percent of these are at speeds of 40 mph or more. By contrast, 23 percent of daily peak gusts from the SSW and SW attain this speed. Although the winter season has a lower average wi nd speed than any other, it also has the greatest frequency of days with peak gusts 40 mph or over (10 percent). This compares with 8 pe rcent for spring, 7 percent for fall, and 5 percent for summer. How ever, reflecting the frequent periods of stagnation in winter, this season also has the highest frequency of days with peak gusts under

Peak wind gust data associ ated with thunderstorm activity are given in 2.3.1.2.1.2.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-8 10 mph (16 percent). This com pares with 10 percent for fall and 1 percent for spring. In summer, such days are virtually non-existent with only one being tabulated in 1,102 days of record. About 60 percent of the days from May through August experie nce drainage winds of at least 13 mph from the west direction for at least two hours daily duri ng the period of 1600 to 2400 PST.

The annual extreme fast est mile of wind speed

• for a given region has been commonly used as the best available measure of wi nd for design purposes (Reference 2.3-9). The standard reference speed level is normally chosen at the 30-ft elevation, and wind speed is assumed to vary with the one-seven th power of height.

All CGS structures have been designed to w ithstand a basic wind (fa stest mile) velocity, including gusts of 100 mph at an eleva tion of 30 ft above the site grade.

This design speed value is conservative for the CGS site since the 100 year return period peak gust as shown in Figure 2.3-4 at HMS is 86 mph at an elevation of 50 ft (as given in Tables 2.3-5 and 2.3-6 at that level peak gusts have not exceeded 80 mp h during the period 1945 to present). The 100 year return period fastest mile of wind would be less than 86 mph since by definition gust velocity divided by an appropriate gust factor provides the velocity of the fastest mile of wind. Although not recorded in histori cal records the 100 y ear fastest mile of wind can be expected to be in the range of 66 to 78 mph. These values are based on the application of gust factors of 1.3 and 1.1 (Reference 2.3-10) for gusts of one and 10 sec durations

• • respectively to the estimated historical value of 86 mph.

2.3.1.2.1.5 High Air Pollution Potential (APP) and Dust Storm Potential. Larson (11) has concluded that "consideration of the general weather paramet ers indicates a significantly high average annual APP over southeastern Washington." Holzworth (12) has estimated that the

mean maximum January mixing depth in the Hanford area is about 250 meters, which is nearly the lowest in the contiguous United States, and for July about 2,000 meters. Hosler (13) has indicated a significantly high frequency of low-leve l inversion in winter o ver this area - on the order of 43 percent with bases below 150 meters. The occurrence of very stable and moderately stable conditions be tween the surface and 60 meters in winter at the Hanford Meteorology Tower is 66.5 percent.

Stagnation is defined by Huschke (14) as "the persistence of a given volume of air over a region, permitting an abnormal buildup of pollutants from sources within the region". Defining the establishment of stagnation as an uninterrupted period of daily average wind speed of 5.0 miles per hour or le ss and/or a peak gust of 15 mile s per hour or less, Jenne (15)

• Fastest mile of wind is generally defined as either the fastest speed associated with 1 mile of passing wind or fastest obser ved 1 minute wind speed.

• • According to Huschke (Glossary of Meteorology, 1959), the duration of a gust is usually less than 20 seconds.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-9 compiled a 15-year summary of Hanford stagnati on periods covering the months of November through February (1947-48 through 1961-62).

Both of the two most notable Hanford stagnation periods experienced during this time occurred in November and December 1952. The first pe riod was from November 15 to December 3 (19 days). Then, after five days of ventilation, stagnation set in again December 9 and lasted through December 28 (20 days). Average wind speeds during the two periods were respectively, 2.6 and 2.9 miles per hour. Eleven days during the first period and eight during the second had peak gusts under 10 miles per hour. One day during the first period and two during the second had average speeds less than 1.0 miles per hour with peak "gusts" of 4 miles per hour. There were 13 day s of fog in each period.

Although stagnation lasting for 20 days can be expected only one season in twenty, a 10-day stagnation period can be expected every other season. Only one season in three will fail to produce a stagnation period of at least eight days.

Air quality in the Hanford area, in terms of sulfur dioxide, nitrogen dioxide, and suspended particulates, is routinely measur ed by the Hanford Environmenta l Health Foundation. (18,29)

For the year 1971, SO2 measur ement in Richland averaged le ss than 0.02 ppm. At other sampling stations, the concentrations were below the detection limit of 0.01 ppm. In 1974, all 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> sequential samples of SO 2 measured in the vicinity of Richland, North Richland, and Hanford 300 Area had concentratio ns below the detection limit of 0.005 ppm which is 25% of the annual average ambient air standard of 0.02 ppm. The 19 7 1 and 1974 measurements for

NO2 and suspended pa r ticu l ates are shown below:

Air Quality Measurements-Annual Averages f o r 1971 and 1974 (18, 2 9* ) (these data are based on 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> integrated samples)

NO2 (pom) Location No. of Samples Max. Min. Avg. Richland (747

Building) 49 6.8 0.06 0.86 Opposite Richland 170 0.019 <.001 0.005 (Hobkirk Ranch) ( 78) (0.022) (0.001) (0.006)

Opposite N.

Richland 157 3.0 <.001 0.024

• Concentrations in parentheses are for 1974.

• • High value due to a local dust storm.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-10 (Gilliam Ranch)

(130) (0.020) (0.001) (0.006) Opposite 300

Area 170 0.025 0.001 0.005 (Sullivan Ranch)

( 77) (0.014) (0.001) (0.005)

Ringold 166 0.028 0.001 0.006 (Keys Ran c h)

White Bluffs 149 0.028 0.001 0.006 (McLane Ranch) Suspe n d ed Particulate

( g/m+3)

Location No. of Samples Max. Min. Avg. Richland (747 Building) 42 (125) 440 (572)** 25 ( 8) 120 ( 57) Opposite Ri c h land - - - -

(Hobkirk Ranch)

Opposite N. Richland

- - - -

(Gilliam Ranch)

Opposite 300 Area - - - - (Sullivan Ranch)

Ringold - - - - (Keys Ran c h)

White Bluffs

- - - -

(McLane Ranch) The major cause of air pollution in the Hanford area is dust occurring during windy periods.

The most significant sources are cultivated fields in the surrounding area. A limited amount of information is available regarding atmospheric dust loading in the Hanford area. Hilst and Nickola (16) conducted limited dust investigations over a range of wind speeds and to heights of 400 feet in the Hanford area. A portion of their findings is presented in Figure 2.3-5. Other investigations which have been made in the Hanford area and reported by Sehmel and Lloyd (17) demonstrate the depe ndence of airborne concentrations on wind speed as shown in Figure 2.3-6.

Measurements of the particulate burden in air at a specific observation point in the 200 Areas at Hanford showed values of around 100 micrograms per cubic meter of air when the wind was less than 8 mph. The particulat e content increased when higher winds were present, averaging 1,000 micrograms per cubic meter with winds of 12 mph, and 3,000 micrograms per cubic meter with winds of 16 mph.

Additional considerations regarding the A ugust 11, 1955 and Januar y 11, 1972 dust storms shown in Figures 2.3-5 and 2.3-6 and other climatological dus t storm characteristics at Hanford are contained in the following paragraphs.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-11 2.3.1.2.1.5.1 Evaluati on of August 11, 1955 and January 11, 1972 Dust Storms at Hanford. The wind speeds at 1.25 ft., 50 ft. and 400 ft.

heights for the A ugust 11, 1955 observation period were 14, 24, 31 mph respectively.

Figure 2.3-5 represents atypical conditions for the site region. The case was or iginally selected for study as a situation with considerable airborne dust conditions com pared to average conditions.

A Hanford climatological summary of dust storms is given in Table 2.3-40 for 1953-1970 (30).

Dust dependence on wind speed and direction (50 ft.) at the Hanford Meteorological Station is given in Table 2.3-41 for the same period (30). Approximat e values of dust concentrations are computed based on an empirical relationshi p using visibility observations (31). The relationship is C6 56v1.25mg/m 3 where V is horizontal visibility in km. This is based on data from the Great Plains with visibilities 7 to 9 miles and wind speeds greater than 12 mph.

Hourly weather observations at the Hanford Meteorological Station were used as input criteria to define a wind resuspension or dust storm period. Hours satisfying dust st orm criteria at Hanford (1953-1970) had either visibilities less than 7 miles and dust reported, or visibilities between 7-14 miles, wind speeds greater than 5.8 m/s, and relative humidities less than 70%.

Since the above empirical concentration - visibility re lationship was based on observ ed dust concentrations at approximately 5-6 feet above the surface, any me asured dust data should be interpolated to that height when comparing the measuremen ts to the calculated 1953-1970 results of Table 2.3-41. (30)

The frequency of the hourly data satisfying the dust storm criteria at Hanford is given in Table 2.3-42. (30)

The August 11, 1955 dust storm has an interpolated 5-ft value of 17 mg/m 3 compared with the climatological average of about 7 mg/m

3. Further inspection of the climatological values in Table 2.3-41 , supports a conclusion that the storm is an example of the more severe type of dust storm that occurs in this region.

Care must be taken in interpretation of Tables 2.3-40 to 2.3-42 to allow for certain limitations. Estimates based on visibilities and/or wind speeds outside the range used in formulation of Equation 1 are of unknown reliab ility. The average visibilities within each wind speed class were within the range of empiri cal validity except for the high wind cases. Another source of errors is the fact that the visibility observat ions are taken at specific times and are not hourly averages. Considering these limitations Tables 2.3-40 to 2.3-42 may be taken to represent overall aspects of the Hanford du st storm climatology. Individual values must be considered approximate estimates - par ticularly those based on only a few data points.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-12 The extreme value in Table 2.3-41 in the 53 to 66 mph class (988 mg/m

3) represents a single observation of 0-1/16 miles visibility for a few minutes. Typica lly at the onset of a dust storm very low visibilities with high wi nds occur for a few minutes.

The very limited visibility and high winds for such a period were coded for th e data for that hour. The station log reveals that five minutes after the ons et of this dust storm, the winds had dropped to 37 mph and visibility was 3/8 mile. The phenomenon was ge nerated by a thunderstorm passing close to the station. Hence, the extreme value in Table 2.3-41 represents an occurre nce of very short duration which was the onset of a dust stor m that had a duration of about one hour. Qualitative observation indicates that this is not an atypical scenario. Over the 40 minute duration of the storm, the average calculated average dust concentration was 60 mg/m

3. It should be noted, however, that th e onset concentration (988 mg/m

3) is of unknown validity because it is calculated from a visibility value for which the empirical model has not been validated.

The visibility during the Januar y 11, 1972 storm was initially less than one mile, changing to four miles during the last half hour of the re ported episode. The January 11, 1972 dust storm had winds at 50' WSW in the range 31 to 43 mph.

Wind storms with peak gusts recorded at 50' on the Hanford Tower during this period have a three to four year return period at Hanford.

Actual particulate loading depends on other factors such as surface conditions and atmospheric stability. Hence, the wind gust return period does not necess arily apply to particulate loading

although it is reasonable to assume the retu rn period is not less than that for wind.

Detailed estimates of the particul ate size and total mass concentr ations cannot be accurately made for the January 11, 1972 dust storm as a result of the lack of any particle size distribution data. In additi on, only one height of mass concentration datum (189 mg/m 3 at 0.2m) was made in the steep gradient region of the vertical profile.

An indication of size distribution and mass loading profile s can be obtained from other data collected at Hanford.

Sehmel (32) reports an April 1972 storm which has mass loadings near the surface which are similar to the January 11, 1972 storm. Althoug h adjacent meteorologica l observations are not available for this episode, the fa ct that the mass loadings at th e lower levels are of the same order as the January 11, 1972 dust storm provides a basis for comparison of the storms. The April 1972 storm has well documented mass loading pr ofiles as a function of particle size. The table shown below contains the profiles of airborne soil concentrations as a function of particle diameter for the dust storm. These are based on optical measurements for smaller particles

(.16 to 5 m) at 0.9m and, impacter-cowl measurements for la rger particles (1 to 230 m) at the indicated heights. The mass loading is dominated by the larger particle data.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-13 Soil Mass Loading for the April 1972 Dust Storm (mg/m

3) Particle Height (m)

Diameter Range ( m) 0.3 1.0 2.0 3.0 10. 32. 0.9 - 5.0 0.21 0.11 *

• 0.058 .0038 5 - 20 0.83 0.28 0.26 .25 0.070 .0056 20 - 60 14. 4.4 2.9 1.5 .81 .29 60 - 240 220. 6.6 2.8 1.3 .19 .11 Total 235 11.4 6.0 ~3.1 1.13 0.41 Comparison of the dust loading of 6

.0 mg/m 3 at 2 meters with the c limatological summaries in Tables 2.3-41 and 2.3-42 indicates that this is a t y pical dust storm for the region.

The particle size distribution for the August 11, 1955 storm is shown in the following table for comparison.

Airborne Dust Loadi ngs of Particles Greater Than 0.

9 For August 11, 1955 Dust Storm (mg/

m 3) Particle Height (m)

Diameter Range ( m) 0.38 2.0 15.2 30.5 0.9 - 5.0 0.015 0.012 0.012 0.0079 5 - 20 0.39 0.32 0.23 0.17 20 - 60 3.1 2.2 1.1 0.77 60 - 240 18. 12.0 2.3 0.78 Total 22. 14.5 3.7 1.7 The April 1972 dust storm has higher mass loadings near the surfa ce in all size ranges. The August 11, 1955 dust storm had hi gher dust loadings above 1 to 2 m heights in the ranges greater than 5 diameter. One interpretation of th ese profiles is that the 1972 storm had a source nearby and the 1955 data re presents advection of airbor ne dust from more remote sources (33).

2.3.1.2.1.5.2 Hanford Dust Storm Climatology for De sign and Operating Bases. The Hanford climatological study of dust storms for 195 3-1970 (30) (discussed in the previous subsection) was re-examined for the purpose of establishing the "worst case" dust storm which may have occurred during that period (33). The worst case dust storm, i.e., that storm which had the largest calculated time in tegrated dust loading (mg-hr/m 3), is considered to be

• No value in reference.

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 2.3-14 160 mg-hr/m 2, duration of 18 hr, and average dust loading of 8.9 mg/m 3 at a height of 5 to 6 ft. The design basis dust st orm is bounded by a postulated volcanic ashfall event (see Section 2.5.1.2.6) in the evaluation of the design and performance of HVAC systems and diesel generators. Results of this worst case dust storm investigation are listed below. As mentioned above, these loadings would apply for a height of 5 to 6 feet above the ground.

Detailed Estimates of the Dust Loadings for the Six Worst Storms Based on Surface Observations of the Hanford Meteorology Station, 1953-1970 Storm Number Total Dust Loading (mg-hr/m 3) Actual Duration (hr) Average Dust Loading (mg/m 3) 1 40* 0.67 60 2 100 1.0 100 3 160 18**

8.9 4 44 2.6 17 5 90 3.1 29 6 80 7 11

The worst storm of these was storm No. 3. While it was also shown in this study that once a given dust storm terminated, there existed a 5% probability that an other one would occur within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> and a 50% probability that another one would occur within 30 days, none of

the above six worst ca se dust storms had occurr ed within 30 days of each other. Most had occurred in different years during the 1953-1970 study period.

The dust loading for storm No. 3 is conservative in terms of its being considered as the worst case storm for use in plant design evaluations. As a result of the shorter storm durations of the measured August 11, 1955, January 11, 1972, and April 1972 dust storms, their time integrated dust loadings at 5-6 feet above the ground are not worse than that computed for storm No. 3 (33).

2.3.1.2.2 Design Snow Load

The American National Standards Institute (ANSI) in "Build ing Code Requirements for minimum Design Loads in Buildings and other Stru ctures" (19) provides weights of 100-year return period ground level snow packs for the site region. The ANSI (Reference 2.3-19) value

• Value is less than actual dust loading as a result of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> duration.

• • The detailed investigation yiel ded 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> as opposed to a range of 1-16 hours given in Table 2.3-40 of 2.3.1.2.1.5.1 for the range in duration of du st storms using the same 1953-1970 data.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1 998 2.3-15 of 20 lb/ft 2 was used as the design snow load for all CGS structures.

• Assuming a specific gravity of 0.1 or snow density 6.24 lb/ft 3 , this design value correspon ds to a snow depth of 3.2 ft. The above snow load is conservative for the site as snow depth seldom exceeds six inches, and the grea test depth of 24.5 inches was reco rded in February 1916. (4) The weight of the 48-hour probable ma ximum winter precipitation can be determined from the data presented in Table 2.3-3. Since the greatest snowfa ll in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was 10.2 inches (February 1993) and a record depth of approximately 12 inches lasted four days (December 1964) these depths would correspond to snow lo ads of 5.3 and 6.24 lb/ft 2 respectively.

2.3.1.2.3 Meteorological Da ta Used for Evaluation of Ultimate Heat Sink

The ultimate heat sink is evaluated in Section 9.2.5.

The meteorological data presented in Figure 2.3-7 and Tables 2.3-1 , 2.3-5 , and 2.3-7a-7h was used to evaluate the perform ance of the CGS spray ponds in 9.2.5 with respect to (1) maximum evaporation and drift loss and (2) minimum water cooling. In accordance with Regulatory Guide 1.27, Rev. 1 "Ultimate Heat Sink fo r Nuclear Power Plants", the worst one-day and 30-day periods of meteorologica l record which resulted in minimum heat transfer to the atmosphere were established. The worst recorded 30-day period (30-day average) of maximum difference between dry-bulb and dewpoint temperature and highe st simultaneously recorded wind speeds which resulted in the maximum evapor ation and drift loss were also established.

Climatological moisture and temperature data pres ented as a function of time of day for each month in Figure 2.3-7, and wind statistics given in Table 2.3-5 were used to establish the maximum initial pond temperature for the ultimate heat sink analyses in 9.2.5. It was determined in

9.2.5 using

these meteorological data, solar radiation formulas contained in ASHRAE Handbook of Fundamentals, and techniques outlined in the John Hopkins University Report "Cooling Water Studies" (E dison Electric Institute Rese arch Report No. 5, Project RP-49, November 1969) that the month of July contained the worst average meteorological data which resulted in the ma ximum initial pond temperature.

The worst day meteorological data was considered to be the given combination of meteorological parameters in a particular consecutive twenty-f our hour period which resulted in the worst pond thermal performance. The following recorded episodes of extreme wet bulb temperatures experienced at the CGS and/or HMS sites were evaluated in 9.2.5 to establish the worst pond thermal performance:

1) August 7-9, 1972 at CGS and HMS, presented in Table 2.3-7a (20),

• Ice loading is included in this CGS estimate.

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 2.3-16 2) July 4-12, 1975 at CGS, presented in Tables 2.3-7b to 2.3-7e from the onsite FSAR meteorological monitoring program,

3) August 4, 1961 at HMS, presented in Table 2.3-7f (21).

The meteorological conditions wh ich occurred on July 10, 1975 at CGS resulted in the worst pond thermal performance as determined in 9.2.5.

The following worst month meteorological data were used in 9.2.5 to establish the second through thirtieth day worst pond thermal performance and worst 30-day drift loss and evaporation (Reference 2.3-21):

a) July 9 - August 8, 1961 at HMS, presented in Table 2.3-7a (minimum heat transfer)

b) July 2 - August 1, 1960 at HMS, presented in Table 2.3-7h (maximum evaporation and drift loss)

Diurnal variations in dry bulb and wet bulb temperatur es for both 30-day periods assumed that the hourly temperature varia tion approximated a sine wa ve of one cycle in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Reference 2.3-21). The average wind speeds during both 30-day periods was approximately 5.5 mph. The root mean squar e average of the hourly wind speed data for the 30-day mass loss period is 6.91 mph.

For conservatism in the therma l analysis, the worst day data for thermal performance was assumed to repeat in the analysis until pond temp erature peaked (three days repetition). For conservatism in the mass loss analysis, an upper bound curve was fit to the drift loss data taken during spray pond testing. The drift loss valu e was obtained from this curve. See 9.2.5 for details.

2.3.2 LOCAL

METEOROLOGY

2.3.2.1 Data Comparisons

The local meteorology prior to CGS plant operation at the CGS site c an be described from FSAR meteorological data procur ed during the period April 1, 1974 to March 31, 1976 from the permanent onsite 7-ft and 2 45-ft meteorological towers. Da ta collected fr om the 245-ft CGS tower had been used for the short-term (accident) and long-term (routine) diffusion estimates. Onsite meteorological data were also obtained from a temporary 23-ft tower which commenced operation in April 1972 for the purpose of determining optimum cooling tower geometric orientation for performance during high wet bulb periods.

The 23-ft meteorological tower data were also used with other regi onal data to establish the potential impact of proposed mechanical draft cooling tower atmos pheric releases in th e vicinity of CGS C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-17 (Reference 2.3-22). The permanent tower data have b een compared where appropriate and possible, with simultaneously recorded and historical data obtaine d from the Hanford Meteorological Station (HMS) for the purpose of documenting the representativeness of the two years of onsite meteorological measurements. For the months of April through August 1974, comparisons have also been made with data from the onsite temporary tower; this tower and instrumentation were dismantled in Se ptember 1974. Monthl y and annual average comparisons between simultaneously recorded and historical data for all the aforementioned meteorological tower sites have indicated that agreement between the data sources is reasonably good.

When comparing sources of data, it should be recognized that at any given time, significant differences can exist between th e reported meteorological conditi ons at the CGS and HMS sites (see, for example, Table 2.3-7a). Differences in the frequencies of occurrence of various meteorological conditions at a given site can also exist from year to year or from one elevation to another elevation at a site for coincident observation times. Any discrepancies between summarized data sources can also be attributed (in addition to s ite separation and instrument height above ground) to differenc es in types and accuracies of instrumentation used and

procedures considered for ac quiring, processing, and analyzi ng raw meteorological data.

Details regarding the onsite meteorological measurement program are presented in 2.3.3.

For the following data comparisons, the following definitions are used:

CGS Data obtained from the permanent 7' and 245' towers at the CGS site; summarized here for April 1, 1974 - March 31, 1976.

CGS (temp). Data obtained from the tempor ary 23' tower at the CGS site; summarized here for April 1, 1974 - August 31, 1975.

HMS: Data obtained from the Hanford Me teorological Station; used here for April 1, 1974 - March 31, 1976.

HMS (hist): Data obtained from the Hanf ord meteorology towe r at the Hanford Meteorological Station, used here for various periods identified in the data comparison listings. The source is AEC Research and Development

Report "Climatography of the Hanford Area", June 1972, BNWL-1605.

Wind Variable: At CGS an hour of data whic h contains less than 15 minutes of any one direction sector; at Hanford, the same but for 20 minutes.

Data obtained at the 33 foot (10 meter) elevation is pres ented since these data have been subsequently used in the site diffusion studies for postulated ground-level release cases.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-18 Wind Calm: At CGS, an hour of data for wh ich the average speed is 0.22 miles per hour or less; at Hanford, average sp eed less than 1 mph (as decided by weather observer, co rresponds to no motion of st rip-chart recorded pen).

Sense of Delta T: Positive values imply rela tive stability, negative values imply relative instability.

The first annual cycle of CGS onsite mete orological data which covered the period April 1, 1974 through March 31, 1975 has been presented in deta il. Local meteorological data collected during the second annual cycle (April 1, 1975 through March 31, 1976) generally portrayed the same characteristic s as indicated by comparison w ith the first annual cycle data. Except for the high wet bulb episode experienced at CGS during July 1975 (refer to 2.3.1.2.3 and 2.3.2.3), no monitored onsite data proved to be more severe in terms of the design and operation of CGS than those data presented in 2.3.1.2. Hence, only the second annual cycle monthly averages have been presented in Table 2.3-8a which summarizes the two years of monitored on-site data with concurrently measured and historic al HMS data. Any significant differences noted between first and second annual cycle onsite dat a and concurrent ly measured CGS and HMS data ar e discussed in 2.3.2.1. Otherwise, conclusions st ated herein for the first annual cycle of data similarly apply to the second annual cycle data. It is observed in Table 2.3-8a that any year to year differences in the summarized monthly mean meteorological data at tend to parallel the differences in the means summarized for the HMS site for corresponding months during the two year monitoring period.

Summaries of joint frequency distributions of wind direction and wind speed by atmospheric stability class and results from accident and rout ine diffusion estimates fo r both annual cycles of CGS onsite meteorological data are presented in subs equent sections.

Magnetic tape files of the two yea rs of hourly onsite data have been transmitted to the NRC.

2.3.2.1.1 Winds

Table 2.3-8b presents monthly and annual CGS joint wi nd speed and direction data for the first annual cycle of monitoring. Similar data for the HMS site are given in Table 2.3-9. The CGS data presented in the above tables were collected at an elevati on of 33 feet above local grade for the one year period of record whereas th e HMS historical data were collected at an elevation of 50 feet above local grade during the period 1955 through 1970 (HMS is approximately 280 feet higher in elevation than the CGS site). Additional wind direction frequency statistics are presented in Table 2.3-10.

The CGS 33 ft and CGS (temp) 23 ft wind direction data given in Table 2.3-10 have similar distributions of direction fr equency and show a bimodal wi nd direction distribution from approximately South and also Northwest. These distributions differ from that given for the HMS site where the direction distribution displays a single peak at approximately West C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-19 Northwest to Northwest. Further, the wind direc tion distribution at the CG S site is much more uniform around the compass than it is at the HMS site. The differences in these distributions may be attributed to the influen ce of terrain features, causing va riability of air flow at the CGS site. This conclusion is stre ngthened by the observation that the CGS monthly wind frequency distributions are similar through-out the period of data acquisition.

Tables 2.3-11 and 2.3-12 provide the 20 longest occurrences of wind direction persistence at CGS for an elevation of 33 feet.

Table 2.3-11 shows persistence in one (22.5 degree) sector while Table 2.3-12 shows persistence within two (45 degrees) adjoini ng sectors; the corresponding stability class distri butions and average wind speed within each stability class are also provided.

It is noted that the majority of the periods of high direction persistence at CGS are associated with unstable, neutral, and mode rately stable atmospheric cond itions and moderate to strong wind speeds. These represent relatively good diffusion conditions.

Table 2.3-12a summarizes the longest persistences of wind direction in one and two sectors at CGS measured during the first and second annual cycles.

The annual frequency and dura tion of episodes of high wind direction persistence at CGS de pend upon the frequency and intens ity of weather systems which result in regional large scale gradient flow.

For example, during th e first annual cycle, the longest persistence in one and two sectors lasted 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> (NW) and 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> (NW, NNW) respectively. During the second annual cycle, th e longest persistence in one and two sectors lasted 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> (NNE) and 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> (N, NNE) respec tively. Wherea s the longest persistence in one sector during the first annual cycle lasted 14 ho urs, the duration of th e first three longest persistences in one sector dur ing the second annual cycle (33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> from NNE, 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> from SSW, and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> from NW) ex ceeded that longe st duration.

Table 2.3-13 presents monthly frequen cy distributions and averages of wind speed measured at the CGS and HMS sites. Considering site separation, elevation of sensors, and instrumentation and pr ocedural differences, the CGS wi nd data appear meteorologically reasonable and demonstr ate consistency among data sources.

2.3.2.1.2 Moisture and Temperature

Diurnal variation and averages of dry-bulb, wet-bulb, and dew-point temperatures for the first annual cycle of monitoring at the CGS and HMS sites are given in Tables 2.3-14 to 2.3-16. Tables 2.3-17 to 2.3-19 present frequency dist ributions of dry-bulb, wet-bulb, and dewpoint temperatures, summarized for the first year of CGS site observations.

Table 2.3-20 contains additional climatological summaries of monthly normals and extreme valu es of temperature and humidity measured at HMS.

Considering the 280 ft difference between the CGS and HMS site s and assuming a dry adiabatic lapse rate of 5.48°F/

1000 ft, one can expect a temperature differ ence of about 1.5°F between the dry-bulb temperature data measured at both sites.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-20 Higher monthly average wet-bulb and dewpoint temperatures occurred at CGS since the CGS site experienced air of slightly higher moisture content than the HMS site. The higher moisture content may be attributed to Columbia River prox imity and irrigation of the fields in the vicinity of CGS. This conclusion is strengthened by the fact that mo isture enhancement at CGS was at a minimum for the months of January, Februar y, and March during the first annual cycle of CGS site observations.

During the second annual cycle of monitoring, it was observed t hat the CGS site experienced air of essentially the sa me moisture content as did the HMS site. The absence of the moisture enhancement at CGS, which was noted during the first annual cycle, may be attributed to reduced evaporation from the proximate river a nd irrigated fields. The periodic occurrences (during the second annual cycle) of cooler dry-bulb temperatures and precipitation deficits when high dry-bulb temperatur es prevailed may have resulte d in reduced evaporation.

2.3.2.1.3 Monthly Precipitation

Diurnal variation of precipitation intensity at CGS and monthly total precipitation at CGS and HMS for the first annual cycle of monitoring are given in Table 2.3-21. Frequency of occurrence of precipitation intensity data from April 1974 through March 1975 at CGS are presented in Table 2.3-22. Frequency of occurrence of wind speed and direction versus precipitation intensity for the sa me year of data is given in Table 2.3-22a. The data show that the CGS site experienced less precipitation than did the HMS site.

The difference can be attributed to site separ ation and the incidence of precipitati on falling in the form of showers of quite limited spatial extent.

The precipitation deficit at CGS may also result from a rain shadow effect from Rattlesnake Mountain. A pr ecipitation gradient is kn own to exist along the slope of this terrain feature.

2.3.2.1.4 Fog

Fog data are unavailable for the site. Although fog has been observed in every month of the year at HMS, it is esse ntially a seasonal phenomenon with 95 percent of it observed during the months of November through February. Inclusion of March and October fog would increase this percentage to 99.7.

Tables 2.3-23 and 24 summarize the duration and persistence statistics for fog occurrences at HMS. Because of the relative proximity of the site to the Columbia River, it is ex pected that the freque ncy of occurrence, inte nsity and duration of fog would be somewhat greater than thes e data indicate (r efer also to 2.3.2.1.5).

Most fog in the Hanford region is of the radi ation type and hence occu rs mostly in conjunction with light wind, inversion or st able atmospheric conditions. The occurrence of fog at the site can therefore be considered as one visual indicator of poor atmospheric diffusion conditions.

Advection and frontal fogs occur occasionally at both HMS and the Tri-City stations of Richland and Pasco. In addition, at Richland and Pasco, there are occasional occurrences of C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-21 steam fog from the Columbia Ri ver. These are not usually deep and many would be classified as ground fog.

Statistics on fog persistence are limited to those at HMS. Although most dense (visibility 1/4 mile or less) fogs do not la st longer than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, a few run for much longer periods as shown in Table 2.3-23. After a period of fog, there freque ntly follows a peri od of atmospheric stagnation with a low stratus overcast and light winds. Such conditions may persist for many days.

2.3.2.1.5 Stability Summaries For the purposes of comparison, the t (temperature difference) and sigma theta (standard deviation of the horizontal wind direction fluctuations) stability classifications which are used in diffusion studies at Hanfor d are given below with the t "Pasquill" and sigma theta classes identified in NRC Regulatory Guide 1.23:

REGULATORY GUIDE 1.23 Pasquill Class t/Z ( F/200 FT)

Sigma* Theta Extremely unstable A Less than -2.1 25.0 Moderately unstable B -2.1 to -1.9 20.0 Slightly unstable C -1.9 to -1.6 15.0

Neutral D -1.6 to -0.6 10.0 Slightly stable E -0.6 to 1.6 5.0 Moderately stable F 1.6 to 4.4 2.5 Extremely stable G greater than 4.4 1.7 HANFORD RESERVATION CLASSIFICATION

Pasquill Class t/Z ( F/200 FT) Sigma Theta Groupings

• • (Degrees) Very unstable Less than -2.5 greater than 22.5 Unstable -2.5 to -1.5 25.5 to 17.5 17.5 to 12.5

• Period of 15 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

• • Note that these sigma theta groupings do not necessarily correspond to any particular Hanford Stability Class on the left, i.e., there can be a maximum of 35 group combinations of t/Z and sigma theta although some combi nations are unlikely to occur.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-22 Neutral -1.5 to 0.5 12.5 to 7.5 7.5 to 3.75 Moderately Stable 0.5 to 3.5 3.75 to 2.1 Very Stable greater than 3.5 less than 2.1 Joint frequency distributions of wind speed and direction by atmo spheric stability class (temperature difference and sigma theta combinations) for the first and second annual cycles of monitoring are presented in Section

2.3.3. Percent

frequency of occurrence of stability (t distribution) at CGS and HMS are given in Table 2.3-25. Although the heights over which t was measured are similar for both sites, it is observed in Table 2.3-25 that CGS experiences air of greater thermal stability than HMS. Th is discrepancy between si te stabilities may be accounted for because of terrain differences. The CGS site se rves as a drainage basin for relatively cool air (especially at night), resulting in strong thermal low-level stratification and the formation of persistent temperature inversi ons. This conclusion is strengthened by the observation that the difference between the sites is much more pronounced during July, August, September and October than during all other months. It is dur ing these months that pooling of relatively cool air is at a maximum due, partly, to mi nimum cloudiness and therefore, enhanced nocturnal cooling occurs at the ground. It is noted in Table 2.3-25 that the percent frequencies of stability types for both annual cycles of monitoring at CGS are very similar.

Frequencies of occurrence of t and sigma theta versus time of day for the first year of onsite meteorological measurements are given in Table 2.3-26 and 2.3-27. Although frequency of occurrence and duration of inversion conditions were not analyzed for the site, stagnation and inversion information contained in 2.3.1.2.1.5 and 2.3.2.1.5 for HMS should be representative for the site (except for the fact that the above data indi cates that CGS experiences a greater frequency of surface-based inversions than does HMS).

Figure 2.3-8 shows probabilities of inversion persistence at HMS from 1952-1969 (2).

2.3.2.2 Potential Influence of the Plant and Its Facilities on Local Meteorology

The shapes and sizes of the buildings erected on the plant s ite will produce a disturbed air flow which alters the initial distribution pattern a nd diffusion rates of plant release airborne contaminants. In the diffusion calcu lations this effect is considered.

Electrical power generation by steam turbines requires dissipation of large quantities of low grade thermal energy. Waste heat produced from the operation of CGS is dissipated by means of six circular mechanical draft cooling towe rs. These evaporative cooling towers release waste heat directly to the atmosphere in the fo rm of sensible and late nt heat. An extended visible plume consisting of li quid water droplets can occur principally during the winter months when periods of cold weather and high relative humidity prevail. Fogging is defined as occurring if visible plumes intersect the ground, buildings, or other elevated structures.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-23 Fog occurs naturally in this re gion, and any cooling to wer fog is an extension of the naturally occurring phenomenon. When air temperatures of 0°C or lower prevail, the additional potential exists for icing on thes e surfaces. At times, small cumulus clouds could form above or remote from the plant, de pending on the atmospheric temperat ure and moisture conditions in the first several thousand feet above the cooling towers. No significant environmental or atmospheric impacts arising from CGS cooling tower operation have been observed or are foreseen based on dispersion meteorological studies performed by Battelle Northwest Laboratories (Reference 2.3-22). Details covering potential environmental impacts arising from CGS cooling tower operation are gi ven in the CGS Environmental Report.

2.3.2.3 Local Meteorological Conditions for Design and Operating Bases

The regional long term meteorologica l conditions provided in Section 2.3.1.2 are applicable for use in establishing the plant design and st ation operating bases.

Except for the high wet bulb episode experienced at CG S during July 1975 (refer to 2.3.1.2.3), none of the local short term meteorological data presented in 2.3.2.1 proved to be more seve re in terms of the design and operation of CGS than those presented in 2.3.1.2. Data collected since January 1, 1984 form the revised long-term design and operating basis for dispersion calculation.

2.3.2.4 Topograp hic Description As shown in Figures 2.1

-1 and 2.1-2 , the plant is located at a grade elev ation of 441 feet MSL in a basin area formed by the Saddle Mountains to the northwest, bluf fs and hills rising to about 900 feet MSL to the north and east, the Horse Heaven Hills to the south and the Rattlesnake Hills and Yakima, Umtanum and Manastach ridges to the west. Topographic cross-sections plotted out to 10 kilometers by sector fr om the plant are given in Figure 2.3-9. Except for the cliffs toward the east across the Columbia River, the region within this circumference is basically flat and featureless and slopes gradually toward the Columbia River. Additional details regarding the regional topography and geology are given in 2.5. The effects of regional topography on local meteorology are discussed in 2.3.1.1 , 2.3.2.1.1 and 2.3.2.1.5. The need to consider plume height relative to land elevations has been obviated by the assumption of a ground-le vel release for the accident and ro utine station release cases which are presented in 2.3.4 and 2.3.5. 2.3.3 ONSITE METEOROLOGICAL MEASUREMENT PROGRAM The permanent onsite meteorol ogical data collection system in use since January 23, 1974 consisted of a 240 ft main tower, an auxiliary seven ft instrument mast, sensors wi th associated C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 9-0 0 0 2.3-24 electronics and recording devices , and a meteorological shelter. A 23 ft onsite temporary tower was also used during the period April 1, 1972 thr ough August 31, 1974.

The Battelle Memorial Institute, Pacific No rthwest Laboratories, had been conducting a continuing two year program of acquisition, pr ocessing, and analysis of meteorological data for Energy Northwest Columbia Generating Sta tion in a contractual arrangement with Burns and Roe, Inc.

The first and second annual cycles of reliable meteorological dat a were collected during the periods April 1, 1974 through Ma rch 31, 1975 and April 1, 1975 through March 31, 1976, respectively. The accuracy of these data had been established primarily through calibrations conducted at quarterly intervals as required through a formal pr ogram of quality assurance. The data were examined for mete orological reasonableness, after corrections were applied per the calibration Reports, through computer edit programs. No data were found to be unreasonable. The annual summari es were compared with th e monthly summaries and all were found to be consistent.

The computer summarization programs (identical for monthly and annual purposes) were tested at quarterly interv als by application to dummy data per the quality assurance program. (23) The computer calculation programs for x/Q were similarly tested. Comparisons between CGS meteorol ogical data and concu rrently measured and historical HMS data have been presented in 2.3.2.1.

2.3.3.1 Permanent Onsite Meteorological Tower and Instru mentation Characteristics

The meteorological tower, which is located approximately 2,500 feet west of the CGS plant site with its base at 455' MSL, cons ists of a 240 ft high primary to wer with a five ft mast extending above it. The primary tower is triangular in shape and of open lattice c onstruction to minimize tower interference with meteorological measurem ents. Wind and temperature measurements were made at the top of the mast and at the 33 ft level. The dew point temperature was measured at the 33 ft level. At the lower le vel the instruments were mounted on an eight ft horizontal boom extending southwes t of the tower. Wind and te mperature measurements were also made at the top of a seven ft mast whic h was located approximatel y 80 feet southwest of the 245 ft tower.

Wind speed measurements were made us ing conventional cup anemometers (Climet Instruments, Model 011-1 Wind Speed Transmitter) which have a response threshold of about 0.6 mph and a distance c onstant of less than five feet. Ov er a calibrated range of 0.6 to 90 mph the accuracy of these in struments is +/- 1% or 0.15 mph (whichever is greater).

Data collection on the 240 foot and seven f oot towers was terminat ed on June 1, 1976 subsequent to the collection of two years of reliable tower data ending March 31, 1976, required for the CGS FSAR. Data collecti on on the Primary and ba ckup towers began on July 1, 1984.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-25 Wind direction measurements were made using lightwe ight vanes (Climet Instruments, Model 012-10 Wind Direction Transmitter).

The response threshold of thes e vanes is about 0.75 mph, and their damp ing ratio and distance constant are approximately 0.4 and 3.3 feet, respectively. Dual potentiometers in the Wind Direction Transmitter produce an electrical signal covering 540° in azimuth with an accuracy of +/-3°. In addition, electroni cs had been included to provid e signals which were proportional to the standard deviation of the wind di rection fluctuations at each level.

Temperature instrumentation had been installed to provide measurements of both the ambient air temperature at the 245, 33 and 7 ft levels, as well as the te mperature differences between these levels. The ambient air temperature and the temperature difference measurement systems were independent of each other to provide for reliability. Atmosphe ric stability delta - T classes were determined solely on the basis of the data from the electronic differencing bridge and not by subtracting the am bient air temperature measur ements. All temperature measurements for both systems were made in aspirated radiation shields (Climet Instruments Model 016-1 or -2) using platinum resistance temperature detectors (Rosemount Engineering Co., Model 104 MB6ABCA). These instruments pr ovided an ambient temperature range from

-30°F to +/-130°F and a temper ature difference range of

+/-15°F. The accuracy of the instruments is +/-.09°F in the measurement of temperatures and +/-0.18°F in the measurement of temperature differences.

The dewpoint temperature was me asured at the 33 ft level of th e tower using a lithium chloride dewpoint sensor (Climet Instru ments, Model 015-12) housed in an aspirate d radiation shield (Climet Instruments Model 016-2). Precipitation was measured at ground level using a tipping bucket rain gage (Meteorology Research Incorporated, Model 302) located about 40 feet west of the main tower. This instrument is accurate to within 1% at rainfall rates up to 3 in./hr and has a resolution of 0.01 in. The instrument bu ilding provided a semi-controlled environment near the tower to house the instrument electroni cs and record the data.

Analog strip chart and digital magnetic tape recorde rs were used to provide redundant data recording capability. The primary data recording system was a seven-track digital magnetic tape recorder (Kennedy, Model 1600) which used 1/2 inch tape. Logarithmically time-averaged wind speed, wind direction, temperature, temperature difference and dewpoint temperature signals were recorded at five minute intervals. The time c onstant of the averaging process was five to ten minutes. The standard deviation of wind dir ection fluctuations during the preceding five minutes at each level and the total precipitation were recorded along with the wind and temperature information. All data, except th e wind direction standard d eviations, were also recorded on strip charts which provided a ba ckup data record to enhance data retrievability.

In addition, since the strip charts contained an essentially inst antaneous record of the signal from each instrument, they provided a rapid means of identifying inst rument malfunctions and were useful in system calibration. These strip charts and magnetic tapes were changed weekly.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 1-0 7 2 2.3-26 In summary, the total system ac curacies for the measured mete orological parameters meet or exceed the following specifications:

air temperature

+/-0.5°C temperature difference +/-0.2°C

humidity (dew point) +/-2.8°C

wind speed 0.5 mph from 0.5 to 5 mph, +/-10% of reading above 5 mph per RG 1.97, Rev. 3.

wind direction

+/-5° These are verified by the e nd-to-end calibrations. Data recovery was better than 90%.

2.3.3.2 Quality Assurance Program

To ensure the quality of the meteorological data collected by the monitoring system, an extensive quality assurance program had been instituted. This program covered all phases of meteorological monitoring from th e initial instrument acquisition th rough the analysis of data.

Periodic checks and calibration of the instrument systems and individual components had been instituted. These periodic chec ks ranged from daily inspection of the strip charts to semiannual calibration of the complete sy stem. All checks, calibrations and maintenance were fully documented, including traceability of test and ca libration equipment to the National Bureau of Standards where necessary. Once collected, the data were protected from loss to the maximum extent possible; the di gital tapes were examined to identify possible instrumentation malfunctions; and the data were then copied onto tw o master tapes. The original weekly tape and one master tape were stored in vaults fo r safekeeping while the second master tape was used in the preparation of data summaries. Finally, to ensure proper operation of computer hardware and software, all computer programs used to summarize or analyze the data were checked quarterly using a standard data input. The computer output from these tests was then saved to document computer operation.

The various phases of the quality assurance prog ram pertaining to the two years of permanent onsite meteorological monitori ng and data processing are di scussed in the following subsections.

2.3.3.2.1 Data Recovery During April 1, 1974 - March 31, 1976

The meteorological data acquisition system wa s put into operation during January 1974. Outages in the digital system precluded an in itiation date for the acquisition of reliable data prior to April 1, 1974 because the processing of an inordinate amount of dat a from strip charts C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.3-27 would have been required. Th e data utilized in the production of the monthly and annual summaries have been obtained exclusively with the primary data recording equipment (magnetic tape digital syst em) since August 1, 1974.

Recourse to data recorded on st rip charts was required at tim es prior to August 1974 to assure, as viewed at the time, a recovery rate of 90% repr esentative data as required by Nuclear Regulatory Guide 1.23. The percentage s of monthly data read from strip charts is listed below:

April 1974 - 23.3% May 1974 - 8.3%

June 1974 - 49.5% (W/S 33' - 59.1%)

July 1974 - 1.6%

From routine and anticipated causes (system maintenance and calibration) modest data losses were experienced on the order of 2% and 1.2% for the first and second annual cycles of onsite data collection respectively. A comparable amo unt, for some of the meteorological quantities, was caused by circumstances b eyond Battelle's control (power outages, delay in receipt of spare parts, etc.). The per centages of missing annual data for various meteorological quantities are listed below:

April 1, 1974-March 31, 1975 April 1, 1975

March 31, 1976 Wind Speed 33' 4.0% 2.7% Wind Direction 33' 4.0% 1.5% Dry Bulb Temperature 33' 2.5% 1.2% Wet Bulb, Dewpoint Temperature 33' 2.7%

1.2% Differential Temperature 33' - 245' 3.1%

1.5%

These percentages are representative of missing data for all meteorological quantities except precipitation. During the first annual cycle of monitoring, data recovery for precipitation was

100%. Precipitation data recovery was co mplete during the second annual cycle of monitoring, except possibly dur ing December 1975 when a sand pl ug was discovered in the rain gauge funnel. As a result, it was estimated that less than 0.1 inch of precipitation was not recorded during that month. For the first annual cycle of monitoring, the data recovery rate was 96% or better for all meteorological quantitie s; this rata was 97% or better during the second annual cycle of monitoring.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-28 2.3.3.2.2 Maintenance and Calibration

Assurance of quality data rests primarily with the calibrations performed at quarterly intervals and reported for July and Octobe r 1974; January, April, July and October 1975; and January and April 1976.

All evidence to date obtained through formal calibrations and routine daily and weekly

inspection had demonstrated that the meteorological system rema ined electronically stable in terms of obtaining data of sufficient quality to meet the requirements in Regulatory Guide 1.23.

The calibration corrections required are tabulated in the response to NRC question 6.4 on the FER. The calibrations established any system inaccuracies by compar isons to standards.

These inaccuracies were corrected by appropriately adjusting the data at the data processing stage as opposed to adjusting the system electronics. The calibrations before and after each calibration period were used to determine if corre ctions were required to account for drift or if offset had occurred. No drift corr ections were required. The offsets were discussed in the FER response. For corrections that were not constant throughout the range of a given parameter, a calibration table or curve was used to correct the data. Calibration corrections were applied as part of the computer programs used to edit and translate the data from the original raw-data tapes to a master f ile of hourly values.

2.3.3.2.3 Data Pro cessing and Analysis

For the two years (1974-1976) of onsite FSAR me teorological monitoring at CGS, all data (magnetic tape and strip chart wher e required) were r un through computer edit programs. No data were found to be unre asonable except for known ca uses as documented in Nonconformance Reports. Data corrections, pe r the Calibration Reports, were applied in the computer programs. Summarization of data ha s been accomplished only at such times as calibration information was available to bracket in time the acquired data.

The data for each hour is represen ted by an average of the data for the last 30 minutes in the hour. The averaging period of 30 minutes was select ed for consistency with

1) the data used to formulate the Hanford Diffusion Model used for routine and accident dose calculations, 2) the recommendations in Regulatory Guide 1.23, and 3) computational economy. The only

exception was wind direction which was averaged over one-hour to facilitate the formulation of wind direction persistence summaries.

One thirty-minute period per hour is consid ered adequate for climatological summaries consisting of averages of many hours. In addition, x/Qs based on thirty-minute averages will be conservative for estimates of the one-hour averages. A ll data products were based on these "hourly" averages.

Wet bulb from the permanent tower was obtained from standard psych rometric formulas presented in the Smithsonian Me teorological Tables, 1971 issue.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-29 The above descriptions relate to data co llected and used in FSAR submittals through Amendment #36. Data collec tion and processing since July 1, 1983 is described in 2.3.3.2.4. The Kennedy Tape Recorder has been replaced with a floppy disk recorder for increased reliability.

In several of the monthly summary reports, the computer programs as applied to dummy data have been compiled as called for in the Quality Assurance Manual (Reference 2.3-23) for the purpose of documenting pr oper programming and proper computer performance.

These computer computations have been verified with hand calc ulations made with the dummy data. The computational programs for x/Q were similarly tested.

2.3.3.2.4 Meteorological Monitoring Program During Plant Operation

The meteorological tower, which is located appr oximately 2500 ft west of the CGS plant site with its base at 455 ft msl consists of a 240-ft high primary tower with a 5-ft mast extending above it. This tower is triangular in shape and of open lattice construction to minimize tower interference with meteorological measurements. Wind and temperature measurements are made at the top of the mast and at the 33-ft elevation by duplicate sets of instruments. One set of instruments is the primary measurement sy stem and the other set constitutes the backup instrumentation. The lower elevation wind speed/direction instruments are mounted on a horizontal boom, extending southwest of the tower.

Wind speed and wind direction measurements are made with a single instrument package that combines a wind speed propeller on the leading (upwind) end of the instrument and a wind direction vane, or tail, on the opposite end. Wind speed meas urement range is 0.5 to 90 mph with a threshold sensitivity of about 1 mph. The wind direction measurements are made by the wind passing over the wind vane portion of the instrument. In addition, electronic modules process the data from these in struments and provide output data which is proportional to the standard deviation of the wind direction fluctuations over 15 minutes.

Temperature instrumentation provi des measurements of the ambi ent air temperature at the 245 and 33-ft elevations. Temperature measurements are made in aspirated radiation shields using platinum RTDs. These instruments provide an ambient temperature range from -50°F to

+150°F. Each set of RTDs (one from 33 ft le vel and one from 245 ft level) are calibrated together in the same temperature bath and electronic modules process the data from these instruments to provid e a temperature difference range of +/-15°F.

The relative humidity is measured at the 33-ft elevation of the tower using an intercap sensor with a range of 0 to 100% RH housed in an aspirated radiation shield. Precipitation is measured at ground level using a tipping bucket rain gauge located about 40-ft west of the main tower. The barometric pr essure is measured by a pressure transmitter located inside the C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-001 2.3-30 Met Tower building. The Met Tower building provides a semi-controlled environment near the tower to house the instrument electronics. Signal conditioning is provided in the Met Tower by two GE FANUC PLC controllers, one for the primary instrumentation and one for the backup instrumentation. The primary controller feeds data to the Supervisory System and the PDIS via the LAN. The backup controller feeds data only to the PDIS via the LAN. Information will be available to all locations for both the primary and backup instruments on the LAN. The backup system does not provide da ta for the barometric pressure or the rain gauge. Wind speed, wind direction, temperature, temperature difference and relative humidity signals are averaged by the c ontrollers using a 15 minute time constant device before sending the information to the control room. In the control room ar e three recorder s which record 245-ft and 33-ft wind speed, wind direction, delta temperature, and ambient temperature at 33-ft elevation. The system accuracies for the measured meteorological parameters are demonstrated to meet or exceed the following specifications by performance of instrument loop calibrations:

Air temperature

+/-0.5°C (+/-0.9°F) Temperature difference +/-0.2°C (+/-0.36°F)

Humidity (dew point) +/-2.8°C (+/-5.0°F)

Wind speed

+/-0.50 mph from 0.5 to 5 mph, +/-10% of reading above 5 mph pe r RG 1.97, Rev. 3.

Wind direction

+/-5.0° This data is processed by the Supervisory System which forms the primary communication vehicle for the meteorological system. The supe rvisory system located at the meteorological tower building and the control room digitizes and multiplexes th e data to the control room where it is restored to analog format and sent to recorders and the PPCRS, as required, on a real-time basis. The data input to the superv isory system is 15-minute average analog values. Longer period averages will also be computed fo r trend analysis and re port generation. These data are routed to satisfy display and proce ssing requirements of the onsite technical support centers (TSC) and the emergency operations facility (EOF). The primary meteorological tower data is stored by the plant data acqui sition system. Instru ment calibrations and maintenance procedures meet the data recovery and system accuracy requirements of

Regulatory Guide 1.23 ex cept as noted above.

The Emergency Dose Projection system provi des redundant data commu nication paths with remote access and redundant power sources as required for routine or emergency preparedness support. The near real time access to both th e primary and backup mete orology systems, via the supervisory system or the LAN, thus sati sfies the emergency prep aredness requirements of Regulatory Guide 1.23.

These two systems are designed to meet or exceed data unavailabi lity statements of Regulatory Guide 1.23. If offsite meteorological data is needed, data can be obtained from a network of meteorological towers located on the Hanford Site using methods described in the Emergency C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-07-044 2.3-31 Preparedness Plan. The accuracy, calibration, and reliability of all data not directly controllable by Energy Northwest is determined by the private/governmental controlling agency. 2.3.3.3 Other Meteorological Measurement Programs Considered for the Data Comparisons 2.3.3.3.1 CGS Temporary Tower A temporary 23-ft onsite tower was used during the period Ap ril 1, 1972, through August 31, 1974, to obtain data input for CG S environmental studies and to provide a comparative overlap with the initiall y measured permanent tower data.

The temporary tower was located in the vicinity of the permanent towers with its base at approximately 448 ft msl. Wind data from the temporary tower were obtained at the 23-ft level while temperature data were acquired at the 3-ft level. Wet bulb data from the temporary tower were establishe d from techniques and dat a contained in the U.S. Department of Commerce, Weather Bureau Office Document, Relative Humidity and Dewpoint Table. As a special quality assurance program was not initiate d for the temporary tower installation, it is not possible to assert that this tower's dat a complied with the requirements contained in Regulatory Guide 1.23.

2.3.3.3.2 Hanford Me teorological Station The Hanford Site maintains a network of mete orological towers, whic h can be accessed for data by telephone or electronic form.

2.3.3.4 Joint Stability - Wind Frequency Summaries Joint frequencies of wind direction and wind speed by atmospheri c stability class (sigma theta by t classes), collected at the 33 ft level of the permanent tower during the period from January 1, 1984 to December 31, 1989 are presented in Table 2.3-28A. The sigma theta/t stability classification approach (see 2.3.2.1.5) has been used by Battelle to maintain consistency with the longer term HMS data to which existing data is compared and to satisfy the data requirements of the Hanford Diffusion Model (HDM) the HDM requires joint measurements of sigma theta and t for the more rest rictive stable diffusi on cases and utilizes the Sutton method with locally derived paramet ers for neutral and unsta ble cases (21). The HDM differs from the standard NRC diffusion m odels as a result of the incorporation of empirically derived diffusion co efficients based on historical experiments performed at Hanford. As a result of the extensive experimental data that were used in deriving the HDM, itis appropriate to consider this model when performing diffusion anal ysis at the Hanford Reservation.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-32 In 2.3.2.1 , comparisons between measured CGS ons ite data and simultaneously recorded and historical HMS data illust rated the following differ ences between sites:

a. The CGS site experienced a small ai r moisture enhancem ent during the first annual cycle of monitoring. During th e second annual cycle, the CGS site experienced air of essentia lly the same moisture cont ent as did the HMS site.

b. The CGS site experiences a biom odal wind direction distribution from approximately south and also northwest.

At HMS, the direction distribution displays a single peak at approximat ely west northwest to northwest.

c. The CGS site experiences air of greater thermal stability than does HMS.

Reasons for these differences were given in 2.3.2.1.

2.3.4 SHORT

TERM DIFFUSION ESTIMATES

2.3.4.1 Objective

In this section brief descriptions of the sources, the receptors, and the methodologies used to calculate the air dispersion factors, /Q, for the Exclusion Ar ea Boundary (EAB), the Low Population Zone (LPZ), and the control room are presented.

2.3.4.2 Exclusion Area Boundary The EAB is located at a distan ce of 1950 m (approxima tely 1.2 miles) from the site. The /Qs were calculated using site-s pecific meteorological data from 1996 to 1999, (Reference 2.3-38). The Joint Frequency Distributions (JFDs), Table 2.3-28 , were used as an i nput to the computer code PAVAN, (Reference 2.3-25) and the /Q results are presented in Table 2.3-37. The /Q values at the EAB are calculated for ea ch hour of data. Th e cumulative probability distribution of these values are determined for each of the wind direction sectors.

Two distributions are calculat ed, Pasquill-Gifford (P-G) with meander sigmas and desert sigmas (includes meander). The distributions represent the annual prob abilities that the given /Q values will be exceeded in each wind direction sector at the exclusion area distance.

Table 2.3-34 incorporates the P-G meander effect and Table 2.3-33 has desert sigmas. From each of the sixteen sector distributions, the /Q value which is exceeded 0.5 percent of the total time was selected.

This value was selected based on th e percentage of total observations rather than the percentage of observations that the wind direction is within the appropriate sector. These 16 sector /Q values are given in Tables 2.3-33 through 2.3-34. The highest of these sixteen values is defi ned as the maximum sector /Q value.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-33 2.3.4.3 Low Population Zone The LPZ is located at a distan ce of 4827 m (approxi mately 3 miles) from the site. The /Qs were calculated using site sp ecific meteorological data fr om 1996 to 1999, (Reference 2.3-38), the JFDs, Table 2.3-28 , were used as an input to the computer code PAVAN, (Reference 2.3-25) and the /Q results are presented in Table 2.3-37. The sector /Q values at the LPZ have been estimat ed for various fixed time intervals of a 30-day period. These time intervals are 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> s, 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 1 - 30 days. The estimates for these time periods are made by interpolation on a log-log plot of the two-hour and annual average values as desc ribed by Regulatory Guide 1.145. These interpolations are carried out for the value wh ich is exceeded 5 per cent of the time, and

0.5 percent

of the time. The in terpolations are displayed in Tables 2.3-35 (Desert) through 2.3-36 (P-G, meander). For these interpolations the 2-hour values are a ssumed equivalent to the 1-hour values. These depicti ons and interpolation schemes ar e identical to those specified in Regulatory Guide 1.145.

2.3.4.4 Control Room The control room air dispersion factors /Q were calculated usi ng the 1996 to 1999 site-specific hourly meteorol ogical data, (Reference 2.3-37). The meteorological data and the input parameters were used as input to the computer code ARCON96, (Reference 2.3-36), and the /Q results are presented in Table 2.3-37. 2.3.4.5 Description of Sources There are 4 sources at CGS that could release radioactivity to the e nvironment following an accident. These sources are described below:

a. The roofline source is a vent (short st ack) on top of the reactor building at a height of 70 m (approximately 229 ft) above the ground through which routine releases take place. Following an accident, the exhaust air from the reactor building passes through the SGT filtrati on system before exiting through the roofline stack. This sour ce is treated as a ground level point source in the /Q calculations.

b. The reactor building railroad bay doors are located at the ground level on the eastside wall of the reactor building. It is assumed that some leakage to the environment takes plac e through these doors.

c. The reactor building walls from the 606 ft level to the 670 ft level (top of reactor building) are made of metal sheets and therefore they are assumed to be C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-34 a diffuse source capable of leaking radioactive materials to the atmosphere, this source is also treated as a ground level release source.

d. The Turbine Building Exhaust System (TBES) is a set of four circular vents (short stacks) located on top of the radwas te building roof. Air from the turbine building is exhausted to the at mosphere through these 4 vents.

2.3.4.6 Control Room Intakes There are three intakes at CGS which draw air into the control room during normal operation as well as post accident. A description of these intakes is given below:

a. Local intake point: The local intake poi nt is a vent located on the west side of the radwaste building wall at an elevation of 527 ft (26.5 m above the ground).

b. Remote intakes: there are two ground level remote intake points which are approximately 180 degrees from each other. One remote intake is located to the north-west side of the turbine building and is labeled remote-1, the other is located to the south-east side of the reactor building and is labeled remote-2.

2.3.4.7 Calculations Formulations for calculating short term /Q values have been de veloped for licensing of nuclear power plants and are described in Regulatory Guide 1.145, (Reference 2.3-26) "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants." For the CGS configuration, it is assumed that accidental releases are made at ground level. This assumption provides a conserva tive estimate of downwind /Q values. NRC code PAVAN (Reference 2.3-25) is used to produce dispersi on estimates with the desert sigma option enabled.

Based on the guidance given in Regulatory Guide 1.145 the /Q values are calculated using three separate equations. The particular e quation which is used de pends upon the existing meteorological conditions. The equations are:

(1) )2A / yz(U 1Q / 10+= (2) ) yz3(U 1Q / 10= (3) yzU 1Q / 10=

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-35 Where /Q is relative concentration (sec/m

3) is 3.14159 U 10 is the wind speed at th e 10 meter level (m/sec) y is the horizontal desert diffusion parameter (m) determined from downwind distance and stability category.

z is the vertical desert di ffusion parameter (m) determined from downwind distance and stability category.

Y represents plume meander and building wake effects (m) and is a function of stability category, wind speed, and downwind distance.

A is the smallest vertical plane cross-sectional area of the reactor building (m 2). During neutral or stable atmos pheric stability conditions, the resu lts of all three equations are used to determine dosages. The values from Equations (1) and (2) are compared and the larger is selected. This value is compared with that computed in Equation 3 and the lower value is selected as the appropriate /Q value.

During all other meteorological conditions (unstable and/or wi nd speeds of 6 m/sec or more), only equations (1) and (2) are considered. The appropriate /Q value is the larger of the two.

Values of Y and z, the horizontal and vertical di ffusion parameters are taken from Regulatory Guide 1.145 for the applicable stability category and downwind distance. For extremely stable condition (Category G), the following equations are applied:

Y(G) = 2/3 Y(F) z(G) = 3/5 z(F) 2.3.5 LONG-TERM (ROUTINE) DIFFUSION ESTIMATES

2.3.5.1 Objectives

The joint wind direction and wind speed by at mospheric stability cla ss data presented in Subsection 2.3.3 was used to assess annual average normalized concentrations, X/Q, for 16 radial sectors extending from the site boundary to a distance of 50 miles from the source.

Tables 2.3-38 provides X/Q and D/Q concentrations fo r a mix mode release assuming desert C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 2.3-36 sigmas no decay, no plume deple tion, recirculation, and a building wake (building height -

70.4 m). D/Q is normalized deposition.

2.3.5.2 Calculations

The calculational techniques used are consistent with the gui dance provided in Regulatory Guide 1.111 "Methods for Estimating Atmos pheric Transport and Disp ersions of Gaseous Effluents in Routine Re leases from Light-Water-Cooled Re actors". The joint frequency data presented in Subsection 2.3.3 were used in conjunction with the following equation to obtain X/Q values at appropriate downwind di stance in each of the 16 sectors.

()()()(,).exp ()./x Q f h x Rxxuxx Dijij e zj fxizj=+2032 2 05 2 2 2 2 12 Where: (,)x Q = average effluent concentration in Ci/m 3 normalized by source release rate (Ci/sec) at distan ce x and direction k.

x = downwind distance from release point.

u i = midpoint value of ith wind speed class.

h e = effective plume height.

zj(x) = vertical standard desert deviation of effluent spread at distance x for the jth stability class.

R fx = factor to account for air recirculation and stagnation.

f ijk = joint probability of the ith wind speed class, jth stability, class, and kth wind direction.

= 3.1416

D = maximum building height of adjacent buildings (D = 70.4 m)

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 2.3-37 The building wake correction Equation 1 must not reduce the X/Q estim ate by more than a factor of 3 or

()2 2 2 0512312xjx Dzjx ()./()/+ Equation 1 assumes a long-term c ontinuous release whose effluent is distributed evenly across a 22.5° sector. The release was assumed to be ground level (i.e., he = 0 in Equation 1). Computer code X0QD0Q, with the Desert si gma option enabled described in NUREG-0324, was used to make two se ts of calculations.

The nearest residences where maximum individual doses with single sector contributions occur at distances of 4.0 miles ENE (Ringold) and 4.2 miles ESE (Taylo r Flats) of CGS. The annual average /Q values for these locations are calculated in Table 2.3-38c , 38f and 39c , 39f. The Columbia bluffs rise to an elevation of 878 ft just south of Tayl or Flat. If it is considered that the low level sigma Z is less than 100 m out to 6 km for the P-G stab ility classes D-G, which are prevalent most of the winters and that low-level winds deflect either north or south near the bluffs (Reference 2.3-2), it is estimated that the total doses for these locations may be once again as large due to cont ributions of favorably oriented wind sectors.

The total dose in the channeling area of the Columbia River should include contributions from four other sectors with deflect ed winds and other channeling effects. The individual doses in that location could be twice as large as for th e single sector, constant wind computations. The drift from the cooling towers shoul d remove some of the effluent and deposit it on the site with the salts and counteract the blu ff effect. The mechanical dra ft cooling towers should entrain part of the effluent, lift it with the plume and thus also make the /Q values over-predictive.

Reasons for these differences were given in Subsection 2.3.2.1.

The results reported by St art and Wendell ("Regional Efflu ent Dispersion Calculation Considering spatial and Temporal Meteorological Variation," Pr eprint volume, Symposium on Atmospheric Diffusion and Air Pollution of the American Meteorological Society, September 9-13, 1974) indicate an average value at these di stances of about 0.65 and a minimum single point value of about 1.75. If these factors are multipli ed by the fraction of plume remaining at the distances in question, about 0.75, to account for the conservatism of the nondepleting model used to arrive at the dose estimates provided in 5.2 of the CGS Environmental Report, it is found that the most critical dose of 9.2 mrem to a child's thyroid (at Taylor Flat) is still within the 10 CFR Part 50 Appendix I de sign objective limit of 15 mrem.

For example, 1.75 x 0.75 x 9.2 = 12.1 mrem. This value would still be conservative because the above recirculation factors do not account fo r the existence of a bluff line immediately downwind of Taylor Flat which will under the mo re stable conditions turn the plume before it reaches Taylor Flat. This effect would further reduce the effective recirculation factor.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 2.3-38 At the nearest point of the ne arest population center, about 9 m iles, the average recirculation factor value from Start and Wendell, 1974, appe ars to be about 0.3 with the maximum single point value about 0.8. In addition to this eff ect, the effect of topographic channeling has been evaluated by conservatively hypothesizing that under stable conditions that winds blowing anywhere from the east to we st (through north) sector might end up in the four sectors containing the majority of the population, SE through WSW (Pas co through Benton City).

Including the effects due to channeling resu lts in an estimated maximum factor of approximately 1.6. Applying the factors for re circulation and fracti on of plume remaining after deposition results in a ma ximum effective factor of (1.6 x 0.8 x 0.67 = 0.86) less than one.

It therefore appears reasonable to conclude that the methodology em ployed to estimate doses is sufficiently conservative for the subject site to ensure t hat the doses to individuals and the general population have not been substantially underestimated due to the inherent assumptions.

2.

3.6 REFERENCES

2.3-1 Baker, D. A., Diffusi on Climatology on the 100-N Area, Hanford Washington, Douglas United Nuclear Company, DUN-7841, Richland, Washington, January 1972.

2.3-2 Stone, W. A., et al.

Climatography of the Hanford Area, Battelle Pacific Northwest Laboratories, BNWL-1605, Richland, Washingt on, June 1972.

2.3-3 Stone, W. A.

Meteorological Instrumentation of the Hanford Area, General Electric, Hanford Atomic Products Operation, HW-62455, Richland, Washington, March 1964.

2.3-4 Phillips, E., Tri-City Area, Kennewi ck-Pasco-Richland, Washington Narrative Climatological Summary, Climatography of the United States No. 20-45, U.S. Department of Commer ce and Economic Development.

2.3-5 Federal Meteorol ogical Handbook No. 1, Surface Observations, U.S. Government Printi ng Office, January 1970.

2.3-6 Fujita, T., Estimate of Maximum Wind Speed of Tornadoes in Three Northwestern States, SMRP Research Pa per No. 92, University of Chicago, December 1970.

2.3-7 Jaech, J. L., Statistical Analysis of Tornado Data for the Three Northwestern States, Jersey Nuclear Company, Ri chland, Washington, December 1970.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 2.3-39 2.3-8 Daubek, H. G., Tornado History and a Discussion of the Tornado Warning System, Battelle Memorial Institute, Pacific Northwest Laboratories Report to Jersey Nuclear Company, Richland, Washington, December 1970.

2.3-9 Thom, H. C. S., "New Distribution of Extreme Winds in the United States,"

Journal for the Structural Division, ASCE, ST-7, July 1968, pp 1787-1801.

2.3-10 ASCE Task Committee Report, Wi nd Forces on Structures, Paper No. 3269, Vol. 126, 1961.

2.3-11 Larson, L. B., Air Pollution Potential Over Southeastern Washington, U.S. Weather Bureau, Walla Walla, Washington, May 1970 (unpublished Presentation).

2.3-12 Holzworth, G. C., "Estimates of Mean Maximu m Depths in the Contiguous United States," Monthly Weather Re view, Vol. 92, pp 235-242, May 1964.

2.3-13 Hosler, C. R., "Low-Level Inversion Frequency in the Contiguous United States," Monthly Weather Review, Vo

l. 89, pp 319-339, September 1961.

2.3-14 Huschke, R. E., ed., Glossary of Terms Frequently Used in Air Pollution, American Meteorological Society, Bo ston, Massachusetts , January 1968.

2.3-15 Jenne, D. E., Frequenc y Analysis of Some Climatological Extremes at Hanford, General Electric, Hanford Atomic Products Operation, Richland, Washington, April 1963.

2.3-16 Hilst, G. R., and Nickola, P. W., "On Wind Erosion of Small Particles,"

Bulletin of the American Meteorolog ical Society, Vol. 40, pp 73-77, February 1959.

2.3-17 Sehmel, G. A., and Lloyd, F. D., Airborne Dust Concentrations, Pacific Northwest Laboratories Annual Report for 1971 to the USAEC Division of Biology and Medicine, Vol.

II, Part 1, Atmospheric Sc iences, Battelle Pacific Northwest Laboratories, BNWL-1651, Richland, Washington, December 1972.

2.3-18 Bramson, P. E., and Corley, J. P.

Environmental Surve illance at Hanford for CY-1971, Battelle Northwest Labor atories, BNWL-1683, Richland, Washington, August 1972.

2.3-19 ANSI A58, 1-1972, "Building Code Requirement s for Minimum Design Loads in Buildings and Other Structures," Am erican National Standards Institute, 1972.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 2.3-40 2.3-20 Simpson, C. L., Computer Printouts of Meteorological Data from the HMS and Hanford No. 2 Sites, March 1-August 31, 1972, BNWL, September 1972.

2.3-21 PSAR WPPSS Nuclear Project No. 1, August 1974 (D ocket No. 50-460).

2.3-22 Droppo, J. G. et al., Atmospheric Effects of Circul ar Mechanical Draft Cooling Towers at Washington P ublic Power Supply System Nuclear Power Plant Number Two, Final Report to Burns and Roe, Inc., November 1976.

2.3-23 Quality Assurance Program for the Acquisition, Processing, and Analysis of Meteorological Data for Washington Public Power Supply System Nuclear Project No. 2, Battelle Pacific Northw est Laboratories, Ri chland, Washington, April 1974.

2.3-24 Regulatory Guide 1.70, Rev. 3, Standard Format and Content of Safety Analysis Reports For Nuclear Power Plants, November 1978.

2.3-25 T. J. Bander, "PAVAN: An Atmos pheric Dispersion Program for Evaluating Design Basis Accidental Rele ases of Radioactive Materials from Nuclear Power Stations," NUREG/CR-2858 (November 1982).

2.3-26 U.S. Nuclear Regulatory Commission, 1979: Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," USNRC Office of Standards Development, Wash., D.C.

2.3-27 Gifford, F. A., 1975:

Atmospheric Dispersion Models for Environmental Pollution Applications. Lectures on Ai r Pollution and Environmental Impact Analyses, American Meteorol ogical Society, pp. 35-38.

2.3-28 PSAR WPPSS Nuclear Pr oject No. 2, February 1973.

2.3-29 Fix, J. J., Enviro nmental Surveillance at Hanf ord for CY-1974, Battelle Northwest Laboratories, BNWL-1910, Richland, Washingt on, April 1975.

2.3-30 Orgill, M. M., G. A. Sehmel , and T. J. Bander, 1974: "Regional Wind Resuspension of Dust," Pacific Northw est Laboratory Annual Report for 1973, to the USAEC Division of Biomedical and Environmental Research, Part 3, Atmospheric Sciences, BN WL-1850, pp. 214-219.

2.3-31 Hagen, L. J. and N.

P. Woodruff, 1973: "Air Polluti on from Duststorms in the Great Plains," Atmos. E nviron., 7, pp. 323-332.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-41 2.3-32 Sehmel, G. A., 1976: "T he Influence of Soil Insertion on Atmospheric Particle Size Distributions," Pacific Northwes t Laboratory Annual Report to ERDA Division of Biomedical a nd Environmental Research, Part 3, Atmospheric Sciences, BNWL-200, pp.99-101.

2.3-33 A. Brandstetter of BN L to J. J. Verderber of Burns and Roe Inc., entitled "Hanford Duststorm Climatology Enviro nmental Studies for CGS," and dated December 2, 1977.

2.3-34 U.S. Atomic Energy Commission, 1972: Regulatory Gu ide 1.23, Onsite Meteorological Programs.

2.3-35 Briggs, G. A., 1973: "Diffusi on Estimation for Small Emissions in Environmental Research Laboratory," Air Resources Atmos. Turbc. and Diffusion Lab. 1973 Annual Repor t, ATDL-106, USDOC-NOAA.

2.3-36 NUREG-6331, "Atmos pheric Relative Concentra tions in Building Wakes,"

Revision 1, May 1997, ARCON96 RSICC Computer Code Collection November CCC-664.

2.3-37 Calculation NE-02-03-14, "Control Room /Q Using ARCON96 with 1996-1999 Meteorological Data."

2.3-38 Calculation N E-02-03-16, "Calculation of the EAB and LPZ /Q Using PAVAN with 1996-1999 Me teorological Data."

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-43 Table 2.3-1 Averages and Extremes of C limatic Elements at Hanford (Based on all Available Records to and Including the Year 1980)

PAGE 1 OF 2 TEMPERATURE (F) DEGREE DAYS (BASE 65F) PRECIPITATION (INCHES) 1912-1980 AVERAGES 1912-1980 EXTREMES DEGREE DAYS (BASE 65F) 1912-1980 TOTALS DAILY MAXIMUM DAILY MINIMUM HEATING 1945-1980 TOTALS COOLING 1960-1980 TOTALS SNOW, ICE PELLETS (SLEET)

DAILY MAXIMUM DAILY MINIMUM MONTHLY HIGHEST MONTHLY YEAR LOWEST MONTHLY YEAR RECORD HIGHEST YEAR RECORD LOWEST YEAR RECORD HIGHEST YEAR RECORD LOWEST YEAR MEAN MONTHLY MAXIMUM MONTHLY YEAR MINIMUM MONTHLY YEAR MEAN MONTHLY MAXIMUM MONTHLY YEAR MINIMUM MONTHLY YEAR MEAN MONTHLY MAXIMUM MONTHLY YEAR MINIMUM MONTHLY YEAR MAX. IN 24 HOURS YEAR MEAN MONTHLY MAXIMUM MONTHLY YEAR MAX. IN 24 HOURS YEAR MAXIMUM DEPTH YEAR Jan Feb Mar Apr May June July Aug Sept Oct Nov Dec

Year 36.6 45.4 56.5 66.2 75.5 83.2 91.8 89.3 79.6 65.4 48.4 39.4 64.8 21.9 27.3 33.7 40.0 47.8 55.3 61.0 59.2 50.8 40.6 31.3 26.0

41.2 29.3 36.3 45.1 53.1 61.7 69.3 76.4 74.3 65.2 53.0 39.8 32.7

53.0 42.5 44.5 49.8 59.6 68.8 75.4 81.8 81.5 71.7 59.0 46.0 38.5

81.8 1953 1958 1926 1934 1947 1922 1960 1967 1967 1952 1954 1957 July 1960 12.1 21.4 39.4 47.5 56.6 63.0 72.4 69.8 58.8 48.8 31.3 18.5

12.1 1950 1929 1955 1955 1933 1953 1963 1964 1926 1930 1955 1919 Jan 1950 72 71 83 95 103 110 115 113 102 90 75 69 115 1971 1924 1960 1934 1936+

1912 1939 1961 1976+

1933 1975 1980 July 1939 -2

-3 24 41 49 55 59 63 52 31 14 -3 -3 1950 1950 1960 1945 1918 1966 1966 1920 1934 1935 1955 1919 Feb 1950+ 53 55 54 60 70 81 82 81 72 60 52 56 82 1971 1932 1942 1956 1956 1924 1925 1961+

1955 1945+

1959+

1975 July 1925 -23

-23 6 12 23 33 39 40 25 6 27 -27 1934 1950 1955 1935 1954 1933 1979 1918 1926 1935 1955 1919 Dec 1919 1104 781 638 381 156 35 4 6 70 376 758 990 5299 1640 1147 794 522 258 90 22 32 179 479 1008 1224 1640 1950 1956 1955 1955 1977 1953 1955 1960 1972 1946 1955 1964 Jan 1950 694 576 476 253 36 3 0

0 10 200 567 822 0 1953 1958 1947 1977 1947 1938 1975+1979+1979+1952 1954 1957 Aug 1979+0 0

0 2 43 183 384 323 102 2 0

0 1039 0 0

0 24 94 310 518 508 216 10 0 0 518 ------

---

---

1977 1971 1969 1960 1967 1967 1971

---

---

July 1960 0 0

0 0

3 57 232 171 27 0 0

0 0 ------

---

---

1978+

1962 1980 1963 1964 1970 1977+

---

---

Apr 1978+ 0.92 0.60 0.37 0.39 0.48 0.54 0.15 0.24 0.31 0.56 0.85 0.89

6.30 2.47 3.08 1.86 1.22 2.03 2.92 0.81 1.36 1.34 2.72 3.05 2.53 3.08 1970 1940 1957 1969 1972 1950 1966 1977 1947 1957 1926 1931 Feb 1940 0.08 T 0 0

0 0

0 0 T 0 T 0.11 0 1977 1967 1942+

1933+

1931 1919 1939+

1955+

1976+

1917+

1976+

1976+

Aug 1955+ 1.08 1.24 0.59 0.58 1.39 1.50 1.25 0.89 0.82 1.91 0.78 1.00 1.91 1948 1916 1949 1980 1972 1934 1942 1977 1947 1957 1966 1958 Oct 1957 5.3 2.3 0.3 T T 0 0

0 0 T 1.4 3.9 13.2 23.4 26.0 4.2 T T 0 0

0 0 1.5 12.7 19.1 26.0 1950 1916 1951 1968+

1960

---

---

---

---

1973 1955 1964 Feb 1916 7.1 18.0 2.2 T T 0 0

0 0 1.5 8.3 5.4 18.0 1954 1916 1957 1968+

1960

---

---

---

---

1973 1978 1965 Feb 1916 12.0 24.5 2.3 0 0

0 0

0 0 1.5 9.1 12.1 24.5 1969 1916 1957

---

---

---

---

---

---

1973 1978 1964 Feb 1916 WIND (mph) RELATIVE HUMIDITY (%) SKY COVER (SCALE 0-10) SOLAR RADIATION (LANGLEYS)*

1945-1980 AVERAGES PEAK GUSTS 1946-1980 AVERAGES 1946-1980 EXTREMES 1946-1980 AVERAGES (SUNRISE TO SUNSET) 1953-1980 AVERAGES DAILY TOTALS 1953-1980 EXTREME DAILY TOTALS PREVAILING DIRECTION MEAN MONTHLY SPEEDS HIGHEST MONTHLY YEAR LOWEST MONTHLY YEAR SPEED DIRECTION YEAR MEAN HIGHEST MONTHLY YEAR LOWEST MONTHLY YEAR HIGHEST YEAR LOWEST YEAR MONTHLY HIGHEST MONTHLY YEAR LOWEST MONTHLY YEAR MONTHLY HIGHEST MONTHLY YEAR LOWEST MONTHLY YEAR HIGHEST MONTHLY YEAR LOWEST MONTHLY YEAR Jan Feb Mar Apr May June July Aug Sept Oct Nov Dec

Year NW NW WNW WNW WNW WNW WNW WNW WNW WNW NW NW

WNW 6.4 7.1 8.5 9.0 8.9 9.2 8.7 8.0 7.5 6.6 6.1 6.1

7.7 10.3 10.8 10.7 11.1 10.5 10.7 9.6 9.1 9.2 9.1 8.2 8.3 11.1 1972 1976 1977+ 1972+ 1965+ 1949 1963 1946 1961 1946 1977 1968 Apr 1972+ 3.1 4.6 5.9 7.4 5.8 7.7 6.8 6.0 5.4 4.4 2.9 3.9

2.9 1955 1963 1958 1958 1957 1950+ 1955 1956 1957 1952 1956 1963+ Nov 1956 80 65 70 73 71 72 69 66 65 63 64 71 80 SW SW SW SSW SSW SW WSW SW SSW SSW SSW SW

SW 1972 1971 1956 1972 1948 1957 1979 1961 1953 1950 1949 1955 Jan 1972 76.4 70.7 55.9 46.9 43.0 39.7 32.2 35.6 41.6 56.8 73.6 80.0 54.3 88.8 86.9 65.9 64.5 61.9 53.5 40.5 47.8 55.5 74.2 88.7 90.5 90.5 1960 1963 1950 1963 1948 1950 1955 1976 1977 1962 1979 1950 Dec 1950 60.0 54.0 44.0 36.9 31.2 30.0 21.9 24.5 33.2 42.5 62.8 69.0 21.9 1963 1967 1965 1966 1966 1949 1959 1967 1974 1952 1976 1968 July 1959 100 100 100 100 100 100 99 100 100 100 100 100 100 1980+ 1980+ 1979+ 1978+ 1978+ 1977+ 1972 1972+ 1978+ 1980+ 1980+ 1980+ Dec 1980+ 13 14 12 9 7 10 6 7 10 10 16 26 6 1963 1962 1965+ 1954 1953 1964+ 1951 1951 1962+ 1952+ 1976+ 1972 July 1951 7.9 7.6 6.8 6.4 5.9 5.3 2.9 3.4 4.1 5.8 7.7 8.1 6.0 9.2 9.3 8.5 8.1 7.7 7.0 4.7 5.9 6.7 8.0 9.1 9.2 9.3 1978 1980 1978 1963 1977+

1950 1976 1968 1978 1975 1972 1962 Feb 1980 4.3 5.9 4.9 3.7 4.5 2.8 0.9 0.6 1.4 3.9 6.2 6.4 0.6 1949 1964 1965 1951 1945 1961 1953 1955 1975 1952 1957 1978 Aug 1955 116 194 335 469 569 627 650 548 415 262 130 89 367 136 238 388 535 634 698 714 613 463 303 180 116 714 1973 1970 1965 1973 1970 1960 1973 1955 1975 1976 1957 1970 July 1973 78 128 293 374 472 537 588 475 326 216 97 57 57 1978 1980 1978 1963 1980 1980 1955 1968 1977 1975 1979+ 1980+ Dec 1980+277 422 542 704 838 821 808 721 591 434 295 196 838 1969 1958 1968 1972 1977 1971 1974 1957 1970 1973 1971 1972 May 1977 16 21 44 75 67 112 118 107 61 33 14 9 9 1976+

1976 1979 1974 1962 1965 1972 1959 1957 1974 1969 1973 Dec 1973 EXTREME AVERAGES OR TOTALS AND YEAR OR SEASON OF OCCURRENCE 1912-1980 TEMPERATURE AVERAGES (F) HIGHEST ANNUAL 56.2 (1958+) LOWEST ANNUAL 50.2 (1929)

HIGHEST WINTER (D-J-F) 41.1 (1933-34)

LOWEST WINTER 24.2 (1948-49)

HIGHEST SPRING (M-A-M) 58.2 (1947)

LOWEST SPRING 48.0 (1955)

HIGHEST SUMMER (J-J-A) 78.2 (1958)

LOWEST SUMMER 70.2 (1980)

HIGHEST FALL (S-O-N) 56.6 (1963) LOWEST FALL 49.5 (1978)

1912-1980 PRECIPITATION TOTALS (IN.)

GREATEST ANNUAL 11.45 (1950)

LEAST ANNUAL 2.99 (1976)

SNOW, ICE PELLETS (SLEET)

GREATEST SEASONAL 43.6 (1915-16)

LEAST SEASONAL 0.3 (1957-58)

1945-1980 WIND SPEED AVERAGE (MPH)

HIGHEST ANNUAL 8.3 (1968+) LOWEST ANNUAL 6.3 (1957)

1945-1980 RELATIVE HUMIDITY AVERAGE (%)

HIGHEST ANNUAL 58.9 (1978) LOWEST ANNUAL 49.4 (1967)

1945-1980 SKY COVER AVERAGES (SUNRISE TO SUNSET, SCALE 0-10)

HIGHEST ANNUAL 6.6 (1978+) LEAST ANNUAL 5.1 (1949)

1953-1980 SOLAR RADIATION AVERAGE DAILY TOTAL (LANGLEYS)

HIGHEST ANNUAL 390 (1973) LOWEST ANNUAL 324 (1980)

• CALORIES/cm 2 +ALSO ON EARLIER YEARS

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-44NUMBER OF DAYS CLEAR (0-3 TENTHS SKY COVER, SR TO SS)

GREATEST ANNUAL (1954-80) 121 1960 LEAST ANNUAL (1954-80) 80 1977

CLOUDY (9-10 TENTHS SKY COVER, SR TO SS)

GREATEST ANNUAL (1954-80) 193 1978 LEAST ANNUAL (1954-80) 85 1966

THUNDERSTORMS GREATEST ANNUAL (1945-80) 23 1948 LEAST ANNUAL (1945-80) 3 1949

HEAVY FOG (VIS. 1/4 MILE OR LESS)

GREATEST SEASONAL (1945-80) 42 1950-51 LEAST SEASONAL (1945-80) 9 1948-49

PRECIPITATION 0.10 INCH OR MORE GREATEST ANNUAL (1946-80) 39 1950 LEAST ANNUAL (1946-80) 10 1965

SNOW 1.0 INCH OR MORE GREATEST SEASONAL (1946-80) 15 1955-56 LEAST SEASONAL (1946-80) 0 1976-77

3 IN. OR MORE SNOW ON GROUND GREASTEST SEASONAL (1946-80) 40 1978-79+

LEAST SEASONAL (1946-80) 0 1976-77

PEAK GUST 40 MPH OR GREATER GREATEST ANNUAL (1945-80) 41 1961 LEAST ANNUAL (1945-80) 10 1978

MAX. TEMPERATURE 90 OR ABOVE GREATEST ANNUAL (1912-80) 85 1940+

LEAST ANNUAL (1912-80) 29 1980

MAX. TEMPERATURE 100 OR ABOVE GREATEST ANNUAL (1912-80) 32 1942 LEAST ANNUAL (1912-80) 1 1954

MAX. TEMPERATURE 32 OR BELOW GREATEST SEASONAL (1912-80) 53 1955-56 LEAST SEASONAL (1912-80) 1 1937-38

MIN. TEMPERATURE 32 OR BELOW GREATEST SEASONAL (1912-80) 141 1916-17 LEAST SEASONAL (1912-80) 75 1957-58

MIN. TEMPERATURE 0 OR BELOW GREATEST SEASONAL (1912-80) 18 1949-50 LEAST SEASONAL (1912-80)01976-77 Table 2.3-1 Averages and Extremes of C limatic Elements at Hanford (Based on all Available Reco rds to and Including the Year 1980) (Continued)

PAGE 2 OF 2 NUMBER OF DAYS (1954-1980) NUMBER OF DAYS (1945-1980)

• CLEAR PTLY CLDY CLOUDY THUNDERSTORMS HEAVY FOG (VIS. 1/4 MI. OR LESS) PRECIPITATION 0.10 INCH OR MORE SNOW 1.0 INCH OR MORE MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MONTHLY MEAN GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR Jan Feb Mar Apr May June July Aug Sept Oct Nov Dec

Year 3 4

6 6

8 10 19 18 15 10 4 3 106 7 9 12 12 14 21 26 30 27 14 10 9 30 1963 1968+

1979+

1962 1973 1961 1960 1955 1975 1980+

1957 1978 Aug 1955 0 0

1 1

1 5 13 9 6

1 1

1 0 1955+

1980+

1978+

1963 1977 1972+

1976+

1978 1978 1975 1973+

1980+

Feb 1980+ 5 5

8 9 11 10 7 7

7 7

5 5 86 22 19 17 15 12 10 5 6

8 14 21 23 172 28 26 23 21 19 15 12 13 16 22 25 28 28 1978 1980+

1977 1979+

1977+

1980+

1976 1968 1977 1973 1973+

1962 Jan 1978+ 17 12 9 6

6 5

0 0

1 9 15 17 0 1963 1964 1979+

1956 1958 1979+

1967 1969+

1975 1970 1961 1978 Aug 1969+ 0

1 2

2 2

2 1

0

1. 10 0 1

1 3

7 8

7 8

4 2

0 1 8 ------

1972+

1969+

1979+

1956 1972+

1975 1953 1959 1976

---

1971 June 1972+ 0 0

0 0

0 0

0 0

0 0

0 0 0 ------

1980+

1980+

1977+

1977+

1963+

1973+

1974+

1976+

1980+

---

1980+

---

6 3

1

0

1 6

8 25 15 11 5 1

1 1

0 1

1 7 13 17 17 1976 1963 1951 1975+

1958 1971

---

1959 1977 1980 1965 1950 Dec 1950 0 0

0 0

0 0

0 0

0 0

0 2 0 1949 1977 1980+

1980+

1980+

1980+

---

1980+

1980+

1978+

1960 1968+ ------ 3 2 2

1 2

2 1

1 1

2 3

3 23 8 5

8 5

4 8

3 4

5 8 10 9 10 1970 1980+

1957 1948 1980+

1950 1974+

1976+

1977+

1950 1973 1964 Nov 1973 0 0

0 0

0 0

0 0

0 0

0 0 0 1977+

1979+

1980+

1977+

1979+

1979+

1980+

1980+

1978+

1978+

1976+

1976+

Aug 1980+ 2 1

0 0

0 0

0 0

1 1

2 6 10 4 2

0 0

0 0

0 0

1 6

6 10 1950 1975+

1957+

---

---

---

---

---

---

1973 1955 1964 Jan 1950 0 0

0 0

0 0

0 0

0 0

0 0 0 1977+

1979+

1980+

---

---

---

---

---

---

1980+

1980+

1976+ ------ NUMBER OF DAYS (1945-1980) NUMBER OF DAYS (1912 - 1980) 3" OR MORE SNOW ON GND. PEAK GUST 40 MPH OR GREATER MAX. TEMP 90 OR ABOVE MAX. TEMP 100 OR ABOVE MAX. TEMP. 32 OR BELOW MIN. TEMP. 32 OR BELOW MIN. TEMP 0 OR BELOW MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLY YEAR MEAN MONTHLY GREATEST MONTHLY YEAR LEAST MONTHLYY YEAR Jan Feb Mar Apr May June July Aug Sept Oct Nov Dec Year 5 3 0

0 0

0 0

0 0

0 1

2 11 23 16 0 0

0 0

0 0

0 0 12 14 23 1969 1950

---

---

---

---

---

---

---

---

1978 1955 Jan 1969 0 0 0

0 0

0 0

0 0

0 0

0 0 1977+

1978+

---

---

---

---

---

---

---

---

1980+

1980+

---

3 2 3

3 2

2 1

1 1

2 2

3 25 11 10 9 7

6 6

4 5

4 8

5 8 11 1972 1976 1956 1972 1971+

1973 1979+

1951 1946 1967 1973+

1957+

Jan 1972 0 0 0

0 0

0 0

0 0

0 0

0 0 1979+

1978+

1978 1979+

1977 1980+

1977+

1980+

1980+

1979+

1979+

1969+

---

0 0 0

3 9 20 18 5

0 0 55 0 0

0 4 11 20 29 29 16 1 0

0 29 ------

---

---

1926 1924 1940+

1941 1915 1938 1933

---

---

July 1941 0 0 0

0 0

0 8

7 0

0 0

0 0 ------

---

---

1980+

1980+

1980+

1963 1948 1977+

1980+

---

---

---

0 0 0

0

2 7

4

0 0

0 13 0 0

0 0

1 9 16 16 2 0

0 0 16 ------

---

---

---

1966+

1970 1971+

1942 1955+

---

---

---

July 1971+ 0 0 0

0 0

0 0

0 0

0 0

0 0 ------

---

---

---

1980+

1980+

1963+

1980+

1980+

---

---

---

---

11 3 #

0 0

0 0

0 0

2 8 24 30 15 2 0

0 0

0 0

0 2 15 19 30 1979 1956 1960

---

---

---

---

---

---

1935 1955 1914 Jan 1979 0 0 0

0 0

0 0

0 0

0 0

0 0 1967+

1976+

1980+

---

---

---

---

---

---

1980+

1976+

1974+

---

27 21 15 5

0 0

0

5 17 25 115 31 28 25 15 3 0

0 0

4 12 30 31 31 1980+

1944+

1944+

1935 1938

---

---

---

1933+

1916 1936 1978+

Jan 1980+ 9 5 6

0 0

0 0

0 0

0 4 14 0 1953 1958 1968+

1974+

1980+

---

---

---

1980+

1962+

1949 1933

---

2 1 0

0 0

0 0

0 0

0

1 4 14 9 0

0 0

0 0

0 0

0 1 14 14 1950 1929

---

---

---

---

---

---

---

---

1955+

1919 Jan 1950+ 0 0 0

0 0

0 0

0 0

0 0

0 0 1977+

1980+

---

---

---

---

---

---

---

---

1980+

1980+

---

REFERENCE NOTES

• PRECIPITATION OBSERVATIONS NOT BEGUN UNTIL 1946

1. LESS THAN 1/2

+ ALSO ON EARLIER YEARS LOCATION AND HISTORY PRESENT LOCATION 25 MILES NW OF RICHLAND, WASHINGTON

LATITUDE 4634'N; LONGITUDE 11936'W, ELEVATION 733 FEET OBSERVATIONS FROM 1912 TO 1944 WERE BY UNITED STATES WEATHER BUREAU COOPERATIVE OBSERVERS AT A SITE ABOUT 10 MILES ENE OF PRESENT LOCATION. SINCE 1944, OBSERVATIONS HAVE BEEN MAINTAINED ON A 24 HOUR-A-DAY BASIS BY THREE DIFFERENT DOE CONTRACTORS.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-45 Table 2.3-2 Average Re t u rn Period (R) and Exis t i ng Record (ER) for Various Precipitation Amounts and Intensity During Specified Time Periods at Hanford (Based on Extreme Value Analys is of 1947-1969 Records)

Amount (Inches)

Intensity (Inches per Hour)

Time Period Time Period R (YEARS) 20 MIN 60 MIN 2 HRS 3 HRS 6 HRS 12 H R S 24 H R S 20 MIN 60 MIN 2 HRS 3 HRS 6 HRS 12 H R S 24 H R S 2 5 10 25 50 100 250 500 1000 0.16 0.24 0.37 0.47 0.53 0.60 0.68 0.73 0.80 0.26 0.40 0.50 0.62 0.72 0.81 0.93 1.02 1.11 0.30 0.48 0.59 0.74 0.85 0.96 1.11 1.22 1.33 0.36 0.55 0.67 0.83 0.96 1.07 1.22 1.33 1.45 0.48 0.77 0.96 1.21 1.40 1.59 1.82 2.00 2.20 0.62 0.95 1.17 1.45 1.66 1.87 2.13 2.34 2.55 0.72 1.06 1.28 1.56 1.77 1.99 2.26 2.47 2.68 0.49 0.72 1.1 1.4 1.6 1.8 2.0 2.2 2.4 0.26 0.40 0.50 0.62 0.72 0.81 0.93 1.02 1.11 0.15 0.24 0.30 0.37 0.42 0.48 0.55 0.61 0.67 0.12 0.18 0.22 0.28 0.32 0.36 0.41 0.44 0.48 0.08 0.13 0.16 0.20 0.23 0.27 0.30 0.33 0.37 0.052 0.079 0.098 0.121 0.138 0.156 0.177 0.195 0.212 0.030 0.044 0.053 0.065 0.074 0.083 0.094 0.103 0.112

• No records have been kept for time peri ods of less than 60 minutes. However, t h e rain gage chart for 6-12-69 shows that 0.55 inch o c curred dur i n g a 20-minute p e riod from 1835 to 1855 PST. An additio nal 0.01 inch occurred between 1855 and 1910 to account for the record 60-m i nute amount of 0.59 inch.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-46 Table 2.3-3 Miscellane o u s Snowfall Statistics (1946 Th r o ugh 1970)

Oct Nov Dec Jan Feb Mar Season Average nu m ber of days w ith depth at 04 0 0 PST 1" o r m o re 3" o r m o re 6" o r m o re 12" o r m o re 0 0

0 0 1 1

0 0 5 2

1

1. 10 5 3

1. 5 3

1 0 #

0 0

0 21 11 5

1. Record grea t est number o f days with depth at 0400 PST 1" or more 3" or more 6" o r m o re 12" o r m o re 0 0 0 0 (1955) 11 (1955) 10 0 0 (1964+) 17 (1955) 14 (1964) 12 (1964) 4 (1969) 31 (1969) 23 (1965) 23 (1969) 1 (1950) 17 (1950) 16 (1969+) 8 0 (1951) 3 0 0

0 (1955-56) 54 (1949-50) 38 (1949-50) 23 (1964-65) 4 Record grea t est depth (1957) 0.3 (1946) 5.1 (1964) 1 2.1 (1969) 1 2.0 (1969) 1 0.0 (1957) 2.3 (Dec 19 6 4) 12.1 Greatest in 24 ho u rs (1957) 0.3 (1955) 4.8 (1965) 5.4 (1954) 7.1 (1959) 5.2 (1957+) 2.2 (Jan. 1954)

7.1 Average

% of water equivalent

of all precipitation 2 14 46 48 29 14 26 ( ) Denotes ye a r of occ u rr e n c e

+ Denotes also in earlier years

1. Denotes less than 1/2 day

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-47 Table 2.3-4 Tornado H i story Within 100 Miles of CGS Date Location June 26, 1916 Wal l a Walla, Washington

April 15, 1925 Condon, O r egon

September 2, 1936 Wal l a Walla, Washington

May 20, 1948 Yak i ma, Washington

May 29, 1948 Yak i ma, Washington

June 11, 1948 Ephr ata, W a shington

June 16, 1948 H a nford R e servation

May 9, 1956 Kenn e wick, Washington

April 12, 1957 Ione, Oregon

April 30, 1957 Yakima, Washington

May 6, 1957 Harr i ngt o n , Washington

April 24, 1958 Wal l a Walla, Washington

June 26, 1958 Wallula Junction, Washington

March 14, 1966 Little Goose Dam, Washington Note: No major damage or loss of life was associated with any of the tornadoes.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-48 Table 2.3-5 Monthly and Annual Prevailing Direction s , Average Speed s , and Peak Gusts: 1945-1970 at HMS (50 ft level)

Peak Gust Month Prev Den Avg Speed Highest Avg Year Lowest Avg Year Speed Den Year Jan NW 6.4 9.6* 1953 3.1 1955 65** S 19 6 7 Feb NW 7.0 9.4 1961 4.6 1963 63 SW 19 6 5 Mar WNW 8.4 10.7 1964 5.9 1958 70 SW 19 5 6 Apr WNW 9.0 11.1 1959 7.4 1958 60 WSW 19 6 9 May WNW 8.8 10.5 1965+ 5.8 1957 71 SSW 19 4 3 June WNW 9.2 10.7 19 4 9 7.7 1950+ 72 SW 19 5 7 July WNW 8.6 9.6 1963 6.8 1955 55 WSW 19 6 3 Aug WNW 8.0 9.1 1946 6.0 1956 66 SW 19 6 1 Sept WNW 7.5 9.2 19 6 1 5.4 1957 65 SSW 19 5 3 Oct WNW 6.7 9.1 1946 4.4 1952 63 SSW 19 5 0 Nov NW 6.2 7.9 1945 2.9 1956 64 SSW 19 4 9 Dec NW 6.0 8.3 1968 3.9 1963+ 71 SW 19 5 5 Year WNW 7.6 8.3 1968+ 6.3 1957 72** SW June 19 5 7

• The average speed for January, 1972, was 10.3 mph

• • On January 11, 1972, a new all-time record peak gu s t of 80 mph was established C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-49 Table 2.3-6 Speed and Direction of Daily Peak Gusts*

Speed Class (mph) Extreme High and Date of Occurrence **

Direction Under 10 10-19 20-2930-39 40-4950-59 60-69 70 orover Total mph Date NNE 0.2 0.8 1.3 0.2 0.1 0 0 0 2.6 47 Feb. 5, 1948 NE 0.3 1.0 1.0 0.2 0 0 0 0 2.5 38 July 10, 1951 ENE 0.2 0.6 0.3 0.1 0 0 0 0 1.2 37 May 27, 1947 E 0.2 0.7 0.2 0.1 # 0 0 0 1.2 44 June 11, 1950 ESE 0.1 0.4 0.1 0 0 0 0 0 0.6 26 June 2, 1958 SE 0.7 2.0 0.4 # # # 0 0 3.1 53 Aug. 29, 1947 SSE 0.7 1.8 0.5 0.1 0.1 # 0 0 3.2 52 Dec. 4, 1952 S 0.3 0.8 1.0 0.7 0.3 0.2 # 0 3.3 58 Dec. 23, 1957 SSW 0.1 0.9 1.5 1.4 0.8 0.4 0.1 # 5.2 71 May 26, 1948 SW 0.2 0.7 3.6 3.4 1.7 0.4 0.1 0.1 10.2 72 June 5, 1957 WSW 0.2 1.5 2.7 2.4 1.1 0.2 0 0 8.1 58 Nov. 3, 1958 L W 0.3 2.2 2.1 1.0 0.3 # 0 0 5.9 52 Nov. 4, 1958 L WNW 1.0 9.6 8.0 5.4 0.6 # 0 0 24.6 50 July 19, 1953 NW 1.5 9.6 6.8 5.1 0.8 0 0 0 23.8 49 April 6, 1952 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-50 Table 2.3-6 Speed and Direction of Daily Peak Gusts* (Continued)

Speed Class (mph) Extreme High and Date of Occurrence **

Direction Under 10 10-19 20-29 30-39 40-49 50-59 60-69 70 or over Total mph Date NNW 0.4 0.8 0.3 # 0 0 0 0 1.5 38 May 8, 1955 N 0.2 1.1 1.4 0.2 0.1 0 0 0 3.0 46 Aug. 27, 1 9 51 Summary 6.6 34.5 31.2 20.3 5.9 1.2 0.2 0.1 100.0 --- ---

• Based on 12 years of observations (1947-58). Tabular values under speed cla s ses denote percent of all daily observations made during the period.

L Denotes the latest of se v e ral occurrences.

1. Denotes less than .05%.

• • A new record was set on January 11, 197 2, when a peak gust of 80 mph was recor d ed at the 50 foot l e vel at the Hanford Meteorology Station. R e ference Document BNWL-1640 "T he Hanford Wind Sto r m of January 11, 1972

." dated February 1972, issued by Battelle P ac i fic Northwest L a boratories.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-51 Table 2.3-7a CGS and HMS Hourly Meteorol ogical Data, August 7-9, 1972 (Ultimate Heat Sink Studies)

CGS Site HMS Tower Site Day/Hour Wind Direction (degrees) Wind Speed (mph) Dry Bulb ( F) Relative Humidity (percent) Wet Bulb ( F) Wind Direction (degrees) Wind Speed (mph) Dry Bulb ( F) Wet Bulb ( F) Elevation 23 feet 23 feet 3 feet 3 feet 3 feet 50 feet 7 feet 7 feet 7 feet 7/1 7/2 7/3 7/4 7/5 7/6 7/7 7/8 7/9 7/10 7/11 7/12 7/13 7/14 7/15 7/16 7/17 7/18 7/19 7/20 7/21 7/22 7/23 7/24 30 50 160 80 50 70 100 70 70 100 130 140 360 80 110 110 90 60 50 50 40 60 100 30 5 5

5 5

5 4

5 4

4 5

5 5

6 8

8 9

9 9 12 12 8 6

5 4 76 72 71 69 65 61 61 59 67 77 85 91 96 99 102 106 107 108 106 106 103 96 89 85 35 36 44 48 51 64 64 68 66 54 46 38 35 32 30 28 25 24 23 23 22 22 26 34 60 58 59 58 55 54 54 53 59 65 69 71 73 74 76 77 77 77 75 75 73 69 66 66 220 270 300 270 180 110 300 320 320 Variable 40 Variable 90 110 90 Variable 90 90 110 90 130 220 300 270 6 6

5 5

4 2

1 5

3 1

3 3

5 5

4 4

8 8

7 8

7 6

8 5 78 76 78 74 72 67 76 81 85 92 93 98 102 104 105 107 108 106 103 97 90 84 82 81 57 56 57 54 55 53 57 60 61 64 64 66 68 68 69 69 69 68 67 64 62 60 58 58 8/1 8/2 8/3 8/4 8/5 8/6 8/7 8/8 8/9 8/10 8/11 8/12 8/13 8/14 8/15 8/16 8/17 8/18 8/19 8/20 8/21 8/22 8/23 8/24 360 130 40 20 40 80 100 140 130 130 120 70 70 40 60 70 30 30 90 110 320 320 320 320 6 5

5 7

6 5

5 4

5 5

7 7

6 6

7 7

8 8

8 8 19 21 15 15 79 72 70 71 71 67 65 62 67 77 85 90 95 98 101 104 106 108 106 106 108 100 95 91 36 39 42 46 46 46 52 54 56 53 46 40 37 34 32 30 28 27 26 25 26 26 28 28 63 58 58 59 59 56 55 53 57 64 69 71 73 74 76 77 77 78 77 76 78 73 71 68 270 270 200 240 200 300 320 20 140 110 110 60 90 90 130 110 220 270 300 300 300 300 300 270 4 4

4 1

4 5

5 3

2 3

4 4

4 4

4 6 10 16 20 21 21 22 16 16 78 77 78 71 70 69 79 85 87 91 93 97 103 104 106 108 108 106 102 96 91 90 87 85 57 57 57 53 53 53 60 63 63 65 65 67 69 69 70 70 70 68 70 66 66 63 64 63 9/1 9/2 9/3 9/4 9/5 9/6 9/7 9/8 9/9 9/10 9/11 9/12 9/13 9/14 9/15 9/16 9/17 9/18 9/19 9/20 9/21 9/22 9/23 9/24 320 320 180 Variable 150 140 160 200 190 200 180 170 190 200 160 300 250 300 310 320 320 320 320 320 17 10 7 5

6 5

5 7

7 8

6 5

7 10 8 13 13 17 14 20 22 17 13 16 90 89 84 77 72 70 71 68 73 81 89 94 98 100 100 102 104 104 104 102 96 93 87 84 32 32 34 35 46 54 57 58 60 56 50 44 38 34 30 30 28 26 27 26 27 29 30 30 69 68 65 61 60 59 61 59 63 68 72 74 76 76 74 76 76 75 76 74 71 70 66 64 270 240 300 240 270 240 240 220 240 320 320 220 220 220 220 240 300 300 300 300 300 300 300 300 9 6

7 6 11 10 8 6 10 9 6

5 10 15 15 13 16 20 20 17 16 18 19 14 83 82 78 80 82 82 82 87 89 94 96 99 101 101 101 103 102 97 93 88 83 80 79 78 63 61 60 62 63 63 63 65 65 65 66 67 68 67 67 68 68 67 65 63 61 60 60 60 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-52 Table 2.3-7b CGS Hourly Meteorological Data, July 4-12, 1975 (33 ft Level) (Ultimate Heat Sink Studies)

Day/Hour Wind Speed mph Dry Bulb F Dewpoint F Wet Bulb F 4/1 4/2 4/3 4/4 4/5 4/6 4/7 4/8 4/9 4/10 4/11 4/12 4/13 4/14 4/15 4/16 4/17 4/18 4/19 4/20 4/21 4/22 4/23 4/24 3.56 3.66 1.71 4.88 2.96 2.76 5.21 3.33 6.02 5.97 12.36 9.67 9.55 8.89 6.28 5.85 6.13 3.55 3.41 5.64 3.88 3.61 3.97 4.81 66.13 64.91 64.16 62.69 61.87 65.60 72.51 77.12 81.71 83.95 90.00 94.48 97.15 99.57 102.37 103.49 104.27 104.77 103.68 95.64 91.49 86.72 83.92 79.60 53.17 54.85 53.92 53.41 53.57 55.20 55.95 54.84 55.52 57.71 55.57 57.09 57.97 58.00 56.77 54.77 52.40 51.20 49.57 59.28 57.55 56.56 58.68 57.84 58.41 58.86 58.07 57.22 56.99 59.30 62.22 63.26 65.16 67.00 67.88 69.99 71.21 71.93 72.17 71.63 70.88 70.56 69.61 71.39 69.30 67.32 67.50 65.66 5/1 5/2 5/3 5/4 5/5 5/6 5/7 5/8 5/9 5/10 5/11 5/12 5/13 5/14 5/15 5/16 5/17 5/18 5/19 5/20 5/21 5/22 5/23 5/24 4.96 3.90 3.30 6.61 6.06 5.00 5.28 2.94 4.87 8.24 5.82 5.69 6.13 4.74 7.52 6.43 4.65 6.93 6.48 6.79 5.25 6.05 3.95 6.34 76.83 74.83 71.68 70.93 71.15 72.99 77.84 81.95 87.01 92.11 96.69 98.96 100.80 103.79 105.36 105.81 104.53 103.48 101.04 96.64 93.63 89.23 87.41 83.80 57.17 55.47 55.41 54.56 55.57 55.89 57.41 58.51 57.89 56.24 54.80 57.68 60.27 54.99 54.77 55.09 53.25 52.80 57.15 57.79 57.23 56.43 57.25 58.84 64.37 62.79 61.64 60.92 61.54 62.36 64.84 66.78 68.07 68.85 69.61 71.60 73.36 71.81 72.18 72.44 71.31 70.80 71.97 70.97 69.80 68.05 67.88 67.55 6/1 6/2 6/3 6/4 6/5 6/6 6/7 6/8 6/9 6/10 6/11 6/12 6/13 6/14 6/15 6/16 6/17 2.68 6.15 4.87 6.08 5.55 2.14 2.35 4.21 2.96 5.10 13.07 12.58 6.00 4.56 5.09 6.27 9.89 81.71 79.55 76.13 73.41 72.83 74.56 80.99 85.01 86.61 89.49 86.75 83.84 87.87 88.69 91.49 94.56 92.59 59.17 59.15 59.73 59.47 59.39 61.28 63.33 61.95 63.65 61.49 62.85 63.47 60.85 61.47 60.59 59.57 60.03 67.05 66.33 65.54 64.49 64.24 65.91 69.16 69.61 71.04 70.70 70.63 70.10 69.87 70.45 70.83 71.21 70.87 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-53 Table 2.3-7c CGS Hourly Meteorological Data, July 4-12, 1975 (33 ft Level)* (Ultimate Heat Sink Studies)

Day/H our Wind Speed mph Dry Bulb F Dewpoint F Wet Bulb F 6/18 6/19 6/20 6/21 6/22 6/23 6/24 7.70 3.03 5.58 12.41 8.22 3.64 3.82 93.44 92.59 90.53 87.60 82.43 79.72 79.49 59.81 59.17 59.57 60.93 58.69 56.40 55.65 71.00 70.43 70.02 69.83 67.03 64.94 64.48 7/1 7/2 7/3 7/4 7/5 7/6 7/7 7/8 7/9 7/10 7/11 7/12 7/13 7/14 7/15 7/16 7/17 7/18 7/19 7/20 7/21 7/22 7/23 7/24 8.39 7.41 8.64 7.31 7.31 8.75 10.86 8.77 10.34 12.46 9.95 10.70 6.33 3.98 8.95 12.70 5.17 3.60 8.61 5.68 4.75 4.06 8.93 16.32 75.23 73.55 72.88 72.67 71.28 74.37 76.77 79.63 83.07 86.32 89.15 90.61 92.80 95.65 95.01 96.85 96.51 97.15 96.40 91.52 86.08 84.35 81.39 81.47 54.69 54.69 53.95 53.68 53.36 55.44 57.23 58.96 59.71 60.35 61.49 60.37 59.71 59.57 58.67 59.31 59.49 58.72 57.76 58.85 61.33 55.89 59.31 60.35 62.52 61.94 61.31 61.10 60.43 62.61 64.38 66.26 67.77 69.13 70.60 70.45 70.76 71.53 70.90 71.76 71.75 71.56 70.88 69.94 69.58 66.22 67.02 67.62 8/1 8/2 8/3 8/4 8/5 8/6 8/7 8/8 8/9 8/10 8/11 8/12 8/13 8/14 8/15 8/16 8/17 8/18 8/19 8/20 8/21 8/22 8/23 8/24 10.89 10.13 11.19 9.25 8.42 8.80 14.06 13.55 M M

M M

M M

M 5.38 5.44 3.66 3.50 6.68 7.28 9.70 9.48 6.22 80.08 79.65 78.05 76.72 74.67 76.45 79.23 80.37 M M

M M

M M

M 101.23 102.03 102.16 100.85 97.09 92.64 89.25 87.49 83.73 58.21 56.69 55.33 54.43 54.53 55.55 56.88 57.60 M M

M M

M M

M 59.09 58.00 56.56 55.28 56.61 56.40 55.25 53.71 54.29 66.01 65.07 63.83 62.91 62.25 63.40 65.03 65.79 M M

M M

M M

M 72.92 72.64 72.02 71.07 70.54 69.09 67.49 66.21 65.24 *M - M i ss i ng data C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-54 Table 2.3-7d CGS Hourly Meteorological Data, July 4-12, 1975 (33 ft Level)* (Ultimate Heat Sink Studies)

Day/Hour Wind S peed m ph Dry Bulb F Dewpoint F Wet Bulb F 9/1 9/2 9/3 9/4 9/5 9/6 9/7 9/8 9/9 9/10 9/11 9/12 9/13 9/14 9/15 9/16 9/17 9/18 9/19 9/20 9/21 9/22 9/23 9/24 7.99 5.26 3.95 4.58 3.78 4.00 4.67 9.88 12.03 10.18 8.76 5.78 7.10 4.77 5.17 5.12 3.24 6.10 6.18 7.80 13.30 25.82 17.04 11.05 81.52 79.12 74.91 73.12 72.37 75.28 79.79 83.38 85.49 88.13 92.13 95.52 99.49 102.61 106.40 108.03 109.47 109.15 107.44 101.47 99.31 94.16 94.61 92.37 55.81 56.45 58.32 58.69 57.92 59.49 60.16 60.06 60.00 61.28 62.85 64.40 63.17 62.72 58.77 55.28 55.49 53.07 52.35 57.87 53.15 61.36 59.49 57.57 65.25 64.76 64.35 63.94 63.25 65.12 66.98 68.06 68.69 70.18 72.20 74.02 74.46 75.01 74.23 73.15 73.65 72.58 71.79 72.42 69.69 72.01 71.19 69.58 10/1 10/2 10/3 10/4 10/5 10/6 10/7 10/8 10/9 10/10 10/11 10/12 10/13 10/14 10/15 10/16 10/17 10/18 10/19 10/20 10/21 10/22 10/23 10/24 8.36 12.16 9.19 5.08 1.56 6.53 7.10 4.15 3.89 5.12 3.77 5.74 5.89 5.27 5.30 5.90 9.37 12.21 8.55 5.54 4.26 3.14 7.36 12.76 90.91 85.92 84.24 80.61 80.24 78.27 83.25 86.77 90.64 92.64 95.23 98.32 100.91 103.09 105.20 105.71 104.93 102.48 101.15 98.27 96.21 90.72 91.33 91.49 58.03 59.17 57.28 56.21 58.48 59.55 62.99 62.91 61.09 62.00 63.36 62.40 59.41 59.39 58.91 56.00 54.11 55.88 56.05 56.13 56.59 60.53 57.68 60.48 69.35 68.39 66.88 65.14 66.21 66.15 69.65 70.67 70.83 71.90 73.38 73.73 72.98 73.58 73.96 72.80 71.78 71.81 71.50 70.68 70.27 70.57 69.31 70.77 11/1 11/2 11/3 11/4 11/5 11/6 11/7 11/8 11/9 11/10 11/11 7.86 12.03 14.22 6.62 8.70 7.13 12.45 M 10.14 12.86 14.68 89.61 86.81 88.56 87.60 85.47 85.28 82.37 M 83.96 83.57 82.51 61.65 62.03 61.48 59.81 60.11 60.37 61.23 M 62.40 62.53 63.97 70.83 70.20 70.42 69.25 68.74 68.82 68.39 M 69.53 69.49 70.00 *M - M i ssing data C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-55 Table 2.3-7e CGS Hourly Meteorological Data, July 4-12, 1975 (33 ft Level)* (Ultimate Heat Sink Studies)

Day/Hour Wind Speed mph Dry Bulb F Dewpoint F Wet Bulb F 11/12 11/13 11/14 11/15 11/16 11/17 11/18 11/19 11/20 11/21 11/22 11/23 11/24 13.48 11.16 11.36 8.17 4.60 4.39 4.27 11.57 9.70 8.24 6.89 10.12 10.69 84.13 87.73 87.76 89.92 94.72 96.19 96.16 89.28 84.99 83.01 82.00 81.97 78.83 64.96 62.91 63.12 61.81 59.97 59.89 58.85 64.05 61.07 60.69 61.09 59.39 57.81 71.06 70.96 71.08 71.00 71.46 71.85 71.33 72.05 69.10 68.28 68.20 67.25 65.38 12/1 12/2 12/3 12/4 12/5 12/6 12/7 12/8 12/9 12/10 12/11 12/12 12/13 12/14 12/15 12/16 12/17 12/18 12/19 12/20 12/21 12/22 12/23 12/24 10.37 7.69 3.17 2.34 5.74 6.75 6.57 4.62 4.23 4.51 5.53 5.53 5.93 6.81 10.87 12.07 15.02 12.21 15.39 13.48 8.45 7.23 9.74 5.68 77.65 76.64 75.87 75.71 73.92 70.51 73.07 76.88 79.71 82.69 84.13 86.43 88.75 90.69 91.31 87.20 86.96 87.39 83.63 81.47 81.76 81.36 79.41 77.73 57.01 56.85 56.32 56.99 58.64 56.51 57.71 59.63 60.93 61.44 61.01 61.65 62.40 62.93 61.73 60.19 62.24 61.36 62.45 60.80 58.77 54.43 44.32 43.97 64.57 64.14 63.59 63.89 64.18 61.81 63.38 65.73 67.38 68.61 68.81 69.86 70.98 71.85 71.37 69.32 70.36 70.00 69.46 67.87 66.86 64.51 59.39 58.63 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-56 Table 2.3-7f 24 Hour HMS Meteorological P r ofile for August 4, 1961 Ho u r Dry Bulb T e mp Wet Bulb Temp Dew Pt Wind (mph)

F F F F 0 1 2 3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 82.0 84.0 86.0 85.0 85.0 85.0 86.0 91.0 92.0 96.0 99.0 103.0 107.0 110.0 112.0 112.0 113.0 110.0 108.0 100.0 98.0 96.0 94.0 93.0 61.0 62.0 63.0 63.0 63.0 62.0 61.0 63.0 63.0 64.0 65.0 67.0 69.0 70.0 71.0 71.0 72.0 70.0 68.0 66.0 66.0 66.0 65.0 64.0 45.0 46.0 48.0 49.0 48.0 46.0 43.0 42.0 42.0 41.0 42.0 44.0 45.0 46.0 48.0 48.0 49.0 45.0 43.0 45.0 45.0 46.0 46.0 45.0 4 5 5 5 5

3 8

7 6

6 7

6 6

5 6

5 5

8 14 19 20 18 16 12 24 Hour Average

Dry Bulb = 96.96 F Wet Bulb = 65.62 F Dew Point = 45.29 F W i nd = 8.37 m p h C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-57 Table 2.3-7g Diurnal Variation in Dry Bulb and Wet Bulb Temperature for Use in Analyzing Second Through Thirtieth Day Pond Thermal Performance (Based On July 9 - August 8, 1961 Hourly Hanford Meteorological Station Data)

Hour Dry Bulb ( F) Wet Bulb ( F) 1 2 3

4 5 6 7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 70.2 68.8 68.3 68.8 70.2 72.5 75.6 79.0 82.8 86.6 90.1 93.1 95.4 96.8 97.3 96.8 95.4 93.1 90.1 86.6 82.8 79.0 75.6 72.5 56.5 56.0 55.8 56.0 56.5 57.3 58.4 59.6 61.0 62.3 63.6 64.7 65.5 66.0 66.2 66.0 65.5 64.7 63.6 62.3 61.0 59.6 58.4 57.3 Daily Average and Variation 82.8 +/- 14.5°F 61.0 +/- 5.2°F

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-58 Table 2.3-7h Diurnal Variation in Dry Bulb and Wet Bulb Temperature for Use in Analyzing First Through Thir tieth Day Maximum Mass Loss (Based On July 2 - August 1, 1960 Hourly Hanford Meteorologi cal Station Data)

Hour Dry Bulb ( F) Wet Bulb ( F) 1 2 3

4 5 6 7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 69.4 67.8 67.3 67.8 69.4 71.9 75.1 78.9 82.9 86.9 90.7 93.9 96.4 98.0 98.5 98.0 96.4 93.9 90.7 86.9 82.9 78.9 75.1 71.9 53.3 52.6 52.4 52.6 53.3 54.3 55.7 57.3 59.0 60.7 62.3 63.7 64.7 65.4 65.6 65.4 64.7 63.7 62.3 60.7 59.0 57.3 55.7 54.3 Daily Average and Variation 82.9 +/- 15.6°F 59.0 +/- 6.6°F

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-59 Table 2.3-8a Summary of CGS Onsite Meteorological Data Collected Du r i ng the First and Second Annual Cycles as Compared to Correspond i ng Hanfo r d Meteorolog i cal Station Data (Historical HMS Data Indi cated for Each Month)

April May June July August September October Site and Sens o r Elevation '74 '75 '74 '75 '74 '75 '74 '75 '74 '75 '74 '75 '74 '75 1. Prevailing W i nd Direction CGS 33'HMS 50' HMS (hist) 50' (1955-197

0)

WNW SSW WNW N/A WNW SSW NW WNW N/A WNW WNW NW WNW N/A WNW S S WNW N/A WNW S S WNW N/A WNW N N NW N/A NW WNW S NW NW WNW 2. Mean Wind S p eed (mph) CGS 33' HMS 50' HMS (hist) 50' (1955-197

0) 9.8 8.0 10.3 9.0 9.0 8.4 8.7 9.0 9.6 8.8 8.5 9.3 9.0 10.5 9.2 7.2 7.6 8.1 8.5 8.6 6.8 7.9 7.5 9.0 8.0 6.5 5.7 7.1 6.8 7.5 4.8 7.2 5.6 7.1 6.7 3. Mean Dry Bu lb Temp. (°F)

CGS 33' HMS 3' HMS (hist) 3' (1950-1970)

52.2 47.6

52.5 48.4 52.5 57.4 59.6

57.9 60.7 61.8 72.5 66.1

73.3 67.3 69.9 73.6 78.7

74.8 80.0 77.5 74.7 70.3

76.3 71.2 75.3 66.9 66.2

68.3 67.9 67.0 51.7 52.1 52.0 52.3 53.2 4. Mean Wet Bu lb Temp. (°F)

CGS 33' HMS 3' HMS (hist) 3' (1950-1970)

44.7 39.7

43.9 40.0 42.8 47.2 48.2

46.5 49.0 49.1 56.0 52.7

54.5 54.0 54.5 57.4 61.5

56.3 62.0 57.9 58.0 55.7

57.0 56.0 57.3 52.6 52.0

52.0 52.0 52.6 43.8 45.3 42.0 45.0 45.4 5. Mean Dew Point Temp. (° F) CGS 33' HMS 3' HMS (hist) 3' (1950-1970)

36.6 29.8

33.3 30.0 30.4 36.6 36.9

34.0 38.6 36.0 43.0 40.8

38.2 42.4 41.2 44.9 50.2

41.0 50.1 42.3 45.6 44.2

43.2 44.6 42.8 39.9 39.5

38.9 38.2 39.5 35.0 38.2 31.0 37.2 36.9 6. Total Precipitation (inches)

CGS HMS HMS (hist) 1946-1970 Mean Total

N/A - Not A v ailable

.55 .53

.46 .42 .44

.44 .47

.28 .38 .50

.06 .46

.12 .14 .66

.45 .09

.71 .32 .16 0.0 1.17 Trace 1.16

.21

.06 0.0

.01 .03 .30

.10 .74

.21 .87 .61 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-60 Table 2.3-8a Summary of CGS Onsite Meteorological Data Collected Du r i ng the First and Second Annual Cycles as Compared to Correspond i ng Hanfo r d Meteorolog i cal Station Data (Historical HMS Data Indicated for Each Month) (Continued)

F i rst A n nual Cycle Sec o nd A n nual Cycle Nove m b er Dece m ber January Febru a ry March Site and S e n s or Elevat i o n '74 '75 '74 '75 '75 '76 '75 '76 '75 '76 April '74-M a rch '75 A p r i l '75- M a rch '76 1. Pre v aili n g Wind Direction CGS 33' H M S 50' H M S (h ist) 5 0' (1 95 5-1 97 0)

S S W S NW NW NW S NW NW NW NW N N W N W NW NW SW NW NW SSW NW NW SW NW N N W S S W W N W N W SW W N W NW NW N/A N/A N W ('55 -'7 0) 2. Mean W i nd Speed (m ph) CGS 33' H M S 50' H M S (h ist) 5 0' (1 95 5-1 97 0) 5.8 7.8 5.5 7.7 6.2 6.4 7.1 5.9 7.2 6.0 6.4 5.0 6.4 4.9 6.4 7.8 1 0.4 7.5 1 0.8 7.0 8.7 9.1 8.9 9.6 8.4 7.2 7.8 9.1 1 0.1 7.6 ('55 - '7 0) 3. Mean Dry B u lb Te m p. (°F) CGS 33' H M S 3' H M S (h i s t) 3' (1 95 0-1 9 7 0) 4 2.1 3 9.5 4 2.1 3 9.3 4 0.1 3 6.8 3 4.2 3 5.7 3 4.5 3 3.4 3 2.3 3 2.4 3 2.0 3 1.5 3 0.3 3 3.8 3 7.7 3 3.6 3 7.3 3 7.5 4 1.9 4 0.8 4 2.0 4 0.6 4 4.0 5 3.1 5 2.1 5 3.4 5 2.6 5 3.5 ('50 - '7 0) 4. Mean Wet Bu l b Te m p. (°F) CGS 33' H M S 3' H M S (h i s t) 3' (1 95 0-1 9 7 0) 3 9.3 3 5.5 3 8.0 3 5.0 3 6.4 3 4.5 3 1.9 3 3.0 3 2.0 3 1.2 3 0.0 3 0.6 3 0.0 3 0.0 2 7.9 3 0.9 3 2.9 3 1.0 3 3.0 3 3.6 3 6.2 3 4.4 3 6.0 3 5.0 3 7.3 4 4.3 4 3.4 4 3.4 4 3.6 4 3.8 ('50 - '7 0) 5. Mean Dew P o i n t Te m p. (°F) CGS 33' H M S 3' H M S (h i s t) 3' (1 95 0-1 9 7 0) 3 6.3 3 0.6 3 3.9 3 0.0 3 1.1 3 1.4 2 8.9 2 9.2 2 8.1 2 7.5 2 6.3 2 8.1 2 6.0 2 7.6 2 3.2 2 6.5 2 5.4 2 5.5 2 5.5 2 7.4 2 7.9 2 4.8 2 6.0 2 5.0 2 7.3 3 5.9 3 4.8 3 3.4 3 4.8 3 3.8 ('50 - '7 0) 6. To t a l Prec i p it a ti on (i n c hes) CGS H M S H M S (h ist) 19 4 6-1 970 Mean T o tal N/A - Not A v a ilable .56 .70

.71 .60 .80 .67 .03 .97 .70 .81 .93 .08 1.43 .5 6 .97 .67 .11 .98 .36 .58 .52 .16 .33 .23 .38 4.92 4.54 6.21 5.87 6.53 ('46 - '7 0)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-61 Table 2.3-8b Frequency of Occurrence of Wind Direction Versus Speed for CGS 33-ft Level (1974-1975)

APRIL SPEED CLASS (MPH) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 5 2 8 15 6 4 0 0 0 0 0 0 0 1 19 22 ENE E 0 0 1 3 2 3 0 0 0 0 0 0 0 0 0 0 3 6 E S E SE 0 0 6 4 4 12 1 6 0 1 0 0 0 0 0 0 11 23 SSE S 0 0 4 8 26 18 21 19 0 12 0 0 0 0 3 7 54 64 SSW SW 0 0 4 4 16 18 29 9 30 15 2 4 0 0 13 6 94 56 WSW W 0 0 3 3 10 14 12 16 5 17 3 4 1 1 7 8 41 63 WNW NW 0 0 7 5 19 23 26 15 40 9 21 11 8 3 2 1 123 67 NNW N 0 0 5 3 17 14 5 5 1 1 0 0 0 0 2 0 30 23 VA R CA L M 0 0 2 0 5 0 0 0 0 0 0 0 0 0 0 0 7 0 UNKNO TOTAL 0 0 0 69 0 224 0 176 0 131 0 45 0 13 14 64 14 720 MAY SPEED CLASS (MPH) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 5 7 11 3 0 3 0 0 0 0 0 0 0 0 16 13 ENE E 0 0 1 8 6 6 2 0 0 0 0 0 0 0 0 0 9 14 E S E SE 0 0 6 1 9 16 0 1 0 0 0 0 0 0 0 0 15 18 SSE S 0 0 9 10 38 27 13 45 0 10 0 0 0 0 0 0 60 92 SSW SW 0 0 5 3 30 15 49 18 16 13 4 3 2 0 0 0 106 52 WSW W 0 0 6 3 19 23 30 35 13 13 1 5 0 0 0 0 69 79 WNW NW 0 0 11 4 24 14 34 10 17 13 10 4 0 2 0 0 96 47 NNW N 0 0 3 4 13 7 7 1 0 0 0 0 0 0 0 0 23 12 VA R CA L M 0 0 7 0 8 0 0 0 0 0 0 0 0 0 0 0 15 0 UNKNO TOTAL 0 0 0 93 0 269 0 248 0 95 0 27 0 4 8 8 8 744 JUNE SPEED CLASS (MPH) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 5 7 12 9 1 4 0 0 0 0 0 0 0 1 18 21 ENE E 0 0 6 3 16 14 9 7 0 0 0 0 0 0 0 0 31 24 ESE SE 0 0 4 4 16 23 6 10 0 0 0 0 0 0 0 0 26 37 SSE S 0 0 7 4 34 20 11 18 0 10 0 2 0 0 0 1 52 55 SSW SW 0 0 6 3 20 11 12 6 12 4 1 5 0 0 0 0 51 29 WSW W 0 0 3 2 15 18 5 14 3 13 1 2 1 3 0 0 28 52 WNW NW 0 0 2 3 24 15 19 20 10 8 10 5 2 2 0 0 67 53 NNW N 0 0 6 6 17 11 3 1 0 0 0 0 0 0 0 0 26 18 VAR CALM 0 0 0 0 4 0 0 0 1 0 0 0 0 0 0 0 5 0 UNKNO TOTAL 0 0 10 81 38 317 26 172 9 70 0 26 0 8 44 46 127 720 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-62 Table 2.3-8b Frequency of Occurrence of Wind Direction Versus Speed for CGS 33-ft Level (1974-1 975) (Continued)

JULY SPEED CLASS (MPH) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 3 11 13 12 2 1 0 0 0 0 0 0 0 0 18 24 ENE E 0 0 3 10 9 14 6 6 0 0 0 0 0 0 0 0 18 30 E S E SE 0 0 6 10 18 26 1 4 0 0 0 0 0 0 0 0 25 40 SSE S 0 0 3 6 37 27 16 32 1 5 0 0 0 0 0 0 57 70 SSW SW 0 0 9 7 16 22 18 14 9 6 2 0 0 1 0 0 54 50 WSW W 0 0 6 7 12 14 11 19 3 9 1 0 2 0 0 0 35 49 WNW NW 0 0 5 11 18 18 18 21 17 13 5 4 0 0 0 0 63 67 NNW N 0 0 10 8 25 22 4 5 2 0 0 0 0 0 0 0 41 35 VAR CA L M 0 0 5 0 24 0 3 0 0 0 0 0 0 0 0 0 32 0 UNKNO TOTAL 0 0 0 120 0 327 0 181 0 65 0 12 0 3 36 36 36 744 AUGU S T SPEED CLASS (MPH) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 16 12 21 19 11 6 0 3 0 2 0 0 0 0 48 42 ENE E 0 0 9 10 6 4 0 4 0 0 0 0 0 0 0 0 15 18 E S E SE 0 0 12 6 9 25 0 1 0 0 0 0 0 0 0 0 21 32 SSE S 0 0 8 7 39 33 16 28 0 4 0 3 0 0 0 0 63 75 SSW SW 0 0 11 8 24 16 17 8 13 0 1 1 0 0 0 0 66 33 WSW W 0 0 9 4 18 13 1 6 0 3 0 0 0 0 0 0 28 26 WNW NW 0 0 8 12 19 27 13 12 22 8 10 8 1 0 0 0 73 67 NNW N 0 0 4 15 35 32 10 10 0 0 0 0 0 0 0 0 49 57 VAR CA L M 0 0 12 0 5 0 1 0 0 0 0 0 0 0 0 0 18 0 UNKNO TOTAL 0 0 0 163 0 345 0 144 0 53 0 25 0 1 13 13 13 744 SEPTE M BER SPEED CLASS (MPH) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 19 21 29 11 10 5 11 2 0 0 0 0 0 0 69 39 ENE E 0 0 20 17 20 7 0 0 0 0 0 0 0 0 0 0 40 24 ESE SE 0 0 15 7 11 11 1 3 0 0 0 0 0 0 0 0 27 21 SSE S 0 0 1 8 13 22 7 25 0 4 0 0 0 0 0 0 21 59 SSW SW 0 0 5 12 18 11 11 3 3 3 1 0 0 0 0 0 38 29 WSW W 0 0 8 12 5 10 3 10 1 4 0 0 0 0 0 0 17 36 WNW NW 0 0 9 9 12 19 17 24 12 8 5 4 1 1 0 0 56 65 NNW N 0 0 12 15 29 38 14 28 3 12 0 0 1 0 0 0 59 93 VAR CALM 0 0 10 0 8 0 1 0 0 0 0 0 0 0 0 0 19 0 UNKNO TOTAL 0 0 0 200 0 274 0 162 0 63 0 10 0 3 8 8 8 720 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-63 Table 2.3-8b Frequency of Occurrence of Wind Direction Versus Speed for CGS 33-ft Level (1974-1 975) (Continued)

O C TOBER SP EE D C L A SS (MP H) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 26 26 15 17 1 0 0 0 0 0 0 0 0 0 42 43 ENE E 0 0 26 20 22 4 1 0 0 0 0 0 0 0 0 0 49 24 E S E SE 0 0 15 15 2 19 0 2 0 0 0 0 0 0 0 0 17 36 SSE S 0 0 16 13 21 25 8 13 0 0 0 0 0 0 0 0 45 51 SSW SW 0 0 15 12 21 13 6 1 0 0 0 0 0 0 0 0 42 26 WSW W 0 0 15 12 11 9 2 10 0 5 0 1 0 0 0 0 28 37 WNW NW 0 0 21 17 11 17 15 12 11 7 6 0 0 0 0 0 64 53 NNW N 0 0 29 37 20 24 9 3 2 0 0 0 0 0 0 0 60 64 VAR CA L M 0 1 16 0 4 0 0 0 0 0 0 0 0 0 0 0 20 1 UNKNO TOTAL 0 1 0 331 0 255 0 83 0 25 0 7 0 0 42 42 42 744 NOVEMBER SP EE D C L A SS (MP H) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 18 10 14 13 4 2 0 0 0 0 0 0 2 1 38 26 ENE E 0 0 13 6 10 2 7 0 0 0 0 0 0 0 1 0 31 8 E S E SE 0 0 12 14 3 13 0 7 0 0 0 0 0 0 0 0 15 34 SSE S 0 0 7 12 28 29 15 29 3 5 0 0 0 0 0 0 53 75 SSW SW 0 0 11 12 32 20 14 6 19 8 1 4 0 0 0 0 77 50 WSW W 0 0 9 12 6 14 5 3 2 1 2 2 0 0 0 3 24 35 WNW NW 0 0 22 27 14 34 7 12 5 0 1 1 0 0 2 1 51 75 NNW N 0 0 24 30 34 17 2 3 0 0 0 0 0 0 0 0 60 50 VAR CA L M 0 0 11 0 2 0 0 0 0 0 0 0 0 0 0 0 13 0 UNKNO TOTAL 0 0 0 250 0 285 0 116 0 43 0 11 0 0 5 15 5 720 DE C E M BER SP EE D C L A SS (MP H) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 12 9 3 5 2 2 0 0 0 0 0 0 0 0 17 16 ENE E 0 0 5 6 4 1 0 0 0 0 0 0 0 0 0 0 9 7 ESE SE 0 0 8 9 1 6 1 5 0 1 0 0 0 0 0 0 10 21 SSE S 0 0 5 11 26 39 25 35 3 14 1 1 0 0 0 0 60 100 SSW SW 0 0 14 14 23 11 29 9 9 2 4 0 3 0 0 0 82 36 WSW W 0 0 16 20 17 15 6 7 3 3 2 3 1 4 0 0 45 52 WNW NW 0 0 27 17 25 59 21 11 6 3 1 0 1 0 0 1 81 91 NNW N 0 0 29 21 27 12 6 0 0 1 0 0 0 0 0 0 62 34 VAR CALM 0 0 8 0 3 0 1 0 0 0 0 0 0 0 0 0 12 0 UNKNO TOTAL 0 0 0 231 0 277 0 160 0 45 0 12 0 9 9 10 9 744 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-64 Table 2.3-8b Frequency of Occurrence of Wind Direction Versus Speed for CGS 33-ft Level (1974-1 975) (Continued)

JANUA RY SP EE D C L A SS (MP H) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 11 13 17 11 6 4 0 0 0 0 0 0 0 0 34 28 ENE E 0 0 10 5 12 10 5 1 0 0 0 0 0 0 6 2 33 18 E S E SE 0 0 15 10 5 14 2 1 0 0 0 1 0 0 1 1 23 27 SSE S 0 0 13 10 17 14 15 16 2 6 1 0 0 0 0 0 48 46 SSW SW 0 0 6 15 18 14 10 5 15 7 3 4 0 2 0 1 52 48 WSW W 0 0 13 8 16 8 4 8 3 4 6 0 3 0 0 0 45 28 WNW NW 0 0 23 20 14 37 13 19 0 3 0 0 0 0 1 10 51 89 NNW N 0 0 29 11 47 17 22 15 0 0 0 0 0 0 4 1 102 44 VAR CA L M 0 0 7 0 3 0 1 0 0 0 0 0 0 0 0 0 11 0 UNKNO TOTAL 0 0 0 219 0 274 0 147 0 40 0 15 0 5 17 44 17 744 FEB R UA R Y SP EE D C L A SS (MP H) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 15 8 15 8 2 0 0 0 0 0 0 0 0 0 32 16 ENE E 0 0 5 4 8 2 0 0 0 0 0 0 0 0 0 0 13 6 E S E SE 0 0 5 6 5 10 1 3 0 1 0 0 0 0 0 0 11 20 SSE S 0 0 14 14 20 11 13 18 4 8 1 2 0 0 0 0 52 53 SSW SW 0 0 9 4 8 9 10 3 9 9 18 4 1 4 0 0 55 33 WSW W 0 0 9 4 7 11 3 6 7 1 4 1 1 1 0 0 31 24 WNW NW 0 0 7 12 14 54 9 45 2 10 2 3 1 2 0 0 35 126 NNW N 0 0 14 16 45 19 24 19 14 1 0 0 0 0 0 0 97 55 VA R CA L M 0 0 5 0 2 0 1 0 0 0 0 0 0 0 0 0 8 0 UNKNO TOTAL 0 0 0 151 0 248 0 157 0 66 0 35 0 10 5 5 5 672 MARCH SP EE D C L A SS (MP H) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UNKNO TOTAL NNE NE 0 0 6 5 8 4 5 1 2 0 0 0 0 0 0 0 21 10 ENE E 0 0 4 6 4 2 1 2 0 0 0 0 0 0 0 0 9 10 ESE SE 0 0 9 2 5 6 1 7 0 3 0 0 0 0 0 0 15 18 SSE S 0 0 5 3 24 28 22 35 3 13 0 0 0 0 0 0 54 79 SSW SW 0 0 1 5 15 9 24 14 16 31 6 14 0 0 0 0 62 73 WSW W 0 0 4 1 7 12 12 2 8 0 6 1 0 0 0 0 37 16 WNW NW 0 0 6 13 21 32 19 27 2 6 0 1 4 4 0 0 52 83 NNW N 0 0 9 11 37 18 23 6 12 9 5 0 0 0 0 0 86 44 VAR CALM 0 0 7 0 5 0 0 0 0 0 0 0 0 0 0 0 12 0 UNKNO TOTAL 0 0 0 97 0 237 0 201 0 105 0 33 0 8 63 63 63 744 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-65 Table 2.3-8b Frequency of Occurrence of Wind Direction Versus Speed for CGS 33-ft Level (1974-1 975) (Continued)

ANNUAL SP E E D C L ASS (MPH) CALM 1-3 4-7 8-12 13-18 19-24 25-UP UN KNO TOT A L NNE NE 0 0 141 131 166 127 50 32 13 5 0 2 0 0 2 3 372 300 ENE E 0 0 103 98 119 69 31 20 0 0 0 0 0 0 7 2 260 189 ESE SE 0 0 113 88 88 181 14 50 0 6 0 1 0 0 1 1 216 327 SSE S 0 0 92 106 323 293 182 313 16 91 3 8 0 0 3 8 619 819 SSW SW 0 0 96 99 241 169 229 96 151 98 43 39 6 7 13 7 779 515 WSW W 0 0 101 88 143 161 94 136 48 73 26 19 9 9 7 11 428 497 WNW NW 0 0 148 150 215 349 211 228 144 88 71 41 18 14 5 13 812 883 NNW N 0 0 174 177 346 231 129 96 34 24 5 0 1 0 6 1 695 529 VA R CALM 0 1 90 0 73 0 8 0 1 0 0 0 0 0 0 0 172 1 UNKNOTOTAL 0 0 10 2005 38 3332 26 1945 9 801 0 258 0 64 264 354 347 8760 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-66 Table 2.3-9 Percentage Frequency D i stribution of 50-ft Wind Direction Versus Speed at HMS (1955-1970)

JANUA RY SPEED CLASS (MPH) FEBRUARY S P EED C LA S S (M PH) DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG S P EED DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG SP E E D N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 2.8 2.0 1.8 1.3 1.8 1.8 2.7 1.6 1.4 1.2 1.4 1.4 1.8 2.4 3.6 3.0 1.3 5.4 1.2 0.8 0.7 0.5 0.6 0.7 1.5 0.8 0.9 1.0 1.4 1.5 2.3 5.6 7.8 2.8 0.1 0.2 0.4 0.2 0.1 0.1 0.2 0.4 0.3 0.4 0.6 1.4 1.7 1.3 5.1 6.6 0.6 0.3 0.1 0.1

0.1 0.2 0.1 0.5 1.0 1.6 0.8 10.6 0.9 1.2 0.1 0.2 0.1 0.1

1. #

0.1 0.2 0.6 0.7 0.4 0.1

0.1

1. #

1. 0.1 0.3 0.3 0.2 #

0.1 0.1

1. 4.7 3.4 2.9 1.9 2.4 2.8 4.8 2.9 3.6 4.8 6.8 6.0 6.1 14.0 19.3 6.5 1.4 5.4 4.3 4.6 4.4 3.0 3.0 3.4 4.0 5.4 7.7 11.5 10.7 8.9 6.3 7.1 6.9 4.2 1.8 N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 2.6 1.8 2.2 1.3 1.3 1.3 1.9 1.2 1.3 1.0 1.0 0.9 1.6 1.9 3.3 2.3 1.3 2.4 1.6 1.3 0.9 0.6 0.7 0.7 1.1 0.8 0.7 1.0 1.6 2.1 3.4 4.9 6.9 2.5 0.2 0.4 0.8 0.2 0.1

0.1 0.3 0.4 0.4 0.8 1.3 2.1 2.9 6.8 6.4 0.8

1. 0.2 0.4 0.2

0.1 0.2 0.3 0.6 1.8 1.4 1.1 1.5 1.5 0.1 #

0.1

0.1 0.5 1.1 0.5 0.2 0.3 0.2

1. #

0.1

0.1 0.2 0.5 0.3 0.1 0.1 0.1

0.1 0.1 0.1

1. 4.8 4.5 3.5 1.9 2.0 2.1 3.4 2.6 2.9 4.2 7.4 7.4 9.3 15.5 18.4 5.7 1.5 2.4 4.5 6.3 3.8 3.3 3.0 3.3 4.0 5.2 6.6 10.5 12.6 10.5 7.8 8.3 7.3 4.8

1.8 TOTAL

38.7 30.1 19.6 7.6 2

.6 0.9 0.2 # 100.0 6.2 TOTAL 30.6 31.0 23.8 9.4 3.0 1.5 0.3 100.0 7.1

1. DENOTES LE S S THAN 0.05%

1. DENOTES LE S S THAN 0.05%

MARCH SPEED CLASS (MPH) APRIL SP E E D CLASS (MPH) DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG S P EED DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG SP E E D N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 1.9 1.4 1.7 0.7 0.9 1.0 1.4 0.6 1.0 0.6 0.8 0.8 1.1 1.1 1.5 1.4 1.3 0.7 1.8 1.6 1.2 0.8 0.8 1.1 1.7 1.1 1.4 1.3 2.0 2.5 3.9 4.4 4.4 2.4 0.2 1.0 0.9 0.3 0.1 0.2 0.2 0.5 0.8 0.7 1.1 2.0 3.0 3.1 5.8 5.3 1.1

1. 0.2 0.6 0.2

0.1 0.1 0.3 0.4 1.2 2.5 2.5 1.2 2.2 1.8 0.2 #

0.1

1. #

0.2 0.7 1.5 1.0 0.2 0.8 0.6

1. #

0.3 0.8 0.4 0.1 0.1 0.1

0.1 0.2 0.2

1. 4.9 4.5 3.5 1.6 1.9 2.4 3.7 2.8 3.7 5.3 9.8 10.4 9.6 14.4 13.7 5.1 1.5 0.7 5.4 6.8 4.8 5.0 4.2 4.4 5.0 7.1 7.4 12.1 13.5 11.7 8.2 9.5 8.9 5.6 2.1 N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 1.6 1.0 1.4 0.6 0.8 0.8 1.0 0.5 0.7 0.7 0.7 0.8 1.1 0.8 1.2 0.9 1.6 0.5 1.6 1.3 1.2 1.0 1.0 1.0 1.5 1.0 1.4 1.5 2.3 2.6 4.2 3.9 3.7 1.9 0.5 0.7 0.6 0.4 0.2 0.5 0.2 0.5 0.8 0.7 1.1 2.1 3.8 4.3 6.0 4.2 0.6 0.3 0.3 0.2 0.1

0.1 0.2 0.3 0.9 2.1 2.5 1.7 4.2 3.2 0.2 #

0.1

0.1 0.4 1.6 1.1 0.5 1.3 1.5

1. #

0.1 0.6 0.4 0.1 0.3 0.4

0.1 0.1

1. # 4.2 3.3 3.2 1.9 2.3 2.0 3.1 2.5 3.2 4.7 9.5 11.3 11.9 16.5 14.2 3.6 2.1 0.5 5.5 6.8 5.3 5.2 5.0 4.5 5.2 6.8 6.7 9.6 12.7 11.4 8.8 11.0 11.0 5.8 2.7 TOTAL 19.9 32.6 26.1 13.5 5

.1 1.8 0.5 100.0 8.6 TOTAL 16.7 31.6 26.7 16.3 6.6 1.9 0.2 100.0 9.1 # DENOTES LESS THAN 0.05%

1. DENOTES LESS THAN 0.05%

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-67 Table 2.3-9 Percentage Frequency D i stribution of 50-ft Wind Direction Versus Speed at HMS (1955-1970) (Continued)

MAY SPEED CLASS (MPH) JUNE SP E E D C L A S S (MP H) DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG S P EED DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG SP E E D N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 1.2 0.6 1.0 0.8 0.9 0.9 1.0 0.6 0.6 0.7 0.6 0.9 1.2 0.8 1.2 1.0 1.9 0.6 1.7 1.4 1.5 1.1 1.2 1.0 1.5 1.3 1.7 1.3 2.4 2.7 4.0 3.9 3.6 1.8 1.3 0.9 0.7 1.0 0.3 0.3 0.3 0.6 0.6 0.5 0.8 1.7 3.5 4.4 6.7 4.9 0.7

1. 0.1 0.3 0.1 0.1

0.1 0.2 0.1 0.4 1.1 1.8 1.6 4.7 4.4 0.1 #

0.1

0.1 0.5 0.7 0.2 1.7 2.6

0.1 0.2

0.2 0.5

1. #

1. 3.9 3.0 3.7 2.3 2.4 2.2 3.2 2.7 2.9 3.3 6.4 9.8 11.4 18.0 17.2 3.6 3.2 0.6 5.7 6.7 6.4 5.0 4.7 4.6 5.4 6.4 5.6 7.4 9.6 10.1 8.3 11.2 12.0 5.5 3.2 N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 0.9 0.6 0.8 0.4 0.5 0.6 0.7 0.5 0.6 0.4 0.7 0.6 0.7 0.6 0.7 0.6 1.6 0.3 2.0 1.8 1.4 1.0 1.3 1.3 1.6 1.1 1.7 1.5 2.5 2.5 4.1 3.5 3.7 2.0 1.7 0.9 0.8 0.8 0.4 0.3 0.3 0.4 0.4 0.3 0.7 2.0 3.5 4.4 6.9 5.1 0.8

1. 0.2 0.4 0.5 0.1

0.1 0.1 0.2 1.0 1.5 1.6 6.2 5.4 0.1 #

0.1

1. 0.1 0.3 0.4 0.3 2.1 3.4

0.1

0.4 0.7

1. #

1. 4.0 3.6 3.6 1.9 2.1 2.2 2.7 2.1 2.7 2.9 6.5 8.6 11.1 19.7 19.0 3.5 3.3 0.3 6.0 7.1 7.5 6.5 5.6 5.0 5.2 5.8 5.5 7.3 8.8 9.7 8.8 12.1 13.0 6.3

3.4 TOTAL

16.5 33.4 27.9 15.1 5.9 1.0 # 100.0 8.7 TOTAL 11.8 34.7 28.0 17

.4 6.7 1.2 100.0 9.3

1. DENOTES LE S S THAN 0.05%

1. DENOTES LE S S THAN 0.05% JULY SPEED CLASS (MPH) AUGU S T SPE E D CLASS (MPH) DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG S P EED DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG SP E E D N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 1.1 0.7 0.9 0.5 0.7 0.7 0.9 0.4 0.7 0.5 0.8 0.7 1.0 0.7 0.9 0.8 2.6 0.4 2.6 2.1 2.1 1.2 1.5 1.3 1.7 1.0 1.5 1.3 2.3 3.0 4.4 4.2 3.8 2.0 1.8 0.6 0.8 0.5 0.2 0.3 0.3 0.3 0.4 0.2 0.5 1.7 2.8 3.5 7.7 5.6 0.7

1. 0.1 0.2 0.2

1. #

0.1 0.2 1.0 1.4 0.9 4.8 5.1 0.1

0.4 0.4 0.1 1.8 3.0

0.1 0.2

0.2 0.5

1. 4.4 3.8 3.7 1.9 2.5 2.3 2.9 1.8 2.5 2.5 6.3 8.5 9.9 19.4 18.9 3.6 4.4 0.4 5.3 6.0 5.7 5.0 4.9 4.9 4.9 5.6 5.3 6.5 9.1 9.5 7.7 11.2 12.4 5.8 3.3 N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 1.3 1.0 1.3 0.7 0.9 0.8 1.1 0.7 0.8 0.6 1.0 0.9 1.3 0.9 1.2 1.0 2.8 0.6 2.5 1.8 1.6 1.1 1.3 1.6 1.8 1.1 1.4 1.6 2.7 3.2 5.1 4.1 3.8 2.3 1.3 0.4 0.5 0.3 0.1 0.2 0.3 0.5 0.6 0.3 0.7 1.6 2.9 4.1 7.6 5.1 0.5 0.1 0.1 0.1

1. #

0.2 0.8 1.4 0.7 4.1 4.5 0.1

1. 0.1 0.2 0.1

1.3 2.3

0.1 0.1 0.2 0.5

1. 4.3 3.4 3.3 1.9 2.4 2.7 3.4 2.4 2.5 3.2 6.4 8.6 11.2 18.2 17.4 3.9 4.1 0.6 5.0 5.1 4.4 4.3 4.5 4.7 4.8 5.5 4.8 6.6 8.1 8.5 7.5 10.6 11.8 5.4

3.0 TOTAL

15.0 37.8 26.1 14.1 5.7 1

.0 100.0 8.6 18.9 38.3 25.7 12.1 4.0 0.9 # 100.0 7.9

1. DENOTES LE S S THAN 0.05%

1. DENOTES LE S S THAN 0.05%

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-68 Table 2.3-9 Percentage Frequency D i stribution of 50-ft Wind Direction Versus Speed at HMS (1955-1970) (Continued)

SEPTE M BER SPEED CLASS (MPH) OCTOBER SP E E D C L A S S (MP H) DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG S P EED DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG S P EED N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 1.9 1.5 1.8 1.3 1.6 1.2 1.4 0.9 0.9 0.8 0.8 1.0 1.3 1.2 1.4 1.3 2.0 1.2 2.4 1.8 1.8 0.9 1.1 1.5 2.1 1.3 1.3 1.6 2.1 3.0 4.9 4.0 3.6 2.6 0.6 0.9 1.0 0.6 0.1 0.2 0.2 0.4 0.5 0.3 0.5 1.3 2.7 3.7 5.6 4.9 0.9 0.3 0.4 0.3 0.1

1. #

0.1 0.1 0.2 1.0 1.3 0.8 2.9 3.3 0.2 #

0.1 0.1

0.1 0.5 0.6 0.3 0.9 1.4

1. #

0.1 0.3 0.1 0.1 0.1 0.2

1. 5.5 4.8 4.6 2.4 2.9 2.9 3.9 2.8 2.6 3.3 6.0 8.7 11.1 14.7 14.8 5.0 2.6 1.2 5.3 6.4 5.6 3.9 3.7 4.1 4.6 5.3 5.2 7.0 9.9 9.4 7.6 9.8 10.5 5.7 2.6 N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 2.6 2.1 2.4 1.4 1.8 1.8 2.5 1.0 1.1 1.0 1.0 1.0 1.7 1.7 2.4 2.1 1.6 2.7 1.7 1.1 0.8 0.7 0.9 1.4 1.9 1.5 1.4 1.4 2.0 2.8 4.6 4.7 4.4 2.5 0.1 0.4 0.3 0.2 0.1 0.1 0.1 0.6 0.7 0.5 0.7 1.5 3.0 3.6 5.0 4.0 0.6

1. 0.1 0.1 0.1

0.1 0.1 0.3 0.7 1.4 1.6 0.7 1.4 1.6 0.1

0.2 0.5 1.0 0.6 0.1 0.4 0.4

0.2 0.4 0.1

0.1

1. 4.8 3.6 3.5 2.2 2.8 3.3 5.1 3.3 3.5 4.5 7.3 9.1 10.7 13.2 12.9 5.3 1.7 2.7 3.9 3.9 3.5 3.6 3.1 3.6 4.2 5.7 6.7 9.7 11.3 9.5 7.1 8.1.

8.0 4.7

1.8 TOTAL

23.5 36.6 23.8 11.0 4.0 0.9 100.0 7.5 TOTAL 31.9 33.9 21.4 8.3 3.2 0.8 # # 100.0 6.7

1. DENOTES LE S S THAN 0.05%

1. DENOTES LE S S THAN 0.05%

NOVEMBER SPEED CLASS (MPH) DECE MBER SPE E D C L ASS (MPH) DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG SP E E D DI R E CTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG S P EED N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 2.7 2.2 1.9 1.8 1.9 2.0 2.5 1.4 1.7 1.4 1.6 1.3 2.1 2.5 3.6 3.0 1.4 4.7 1.3 0.7 0.5 0.4 0.6 1.0 1.4 1.2 1.2 1.2 1.6 2.1 3.4 4.6 5.9 2.8 0.1 0.4 0.4 0.1

0.1 0.3 0.4 0.5 0.8 1.4 1.9 1.9 4.9 4.7 1.0 0.2 0.2

0.1 0.2 0.5 0.9 1.5 1.3 0.6 1.0 0.7 0.2 #

0.1 0.3 0.7 0.8 0.4 0.1 0.3 0.1

1. #

0.2 0.3 0.1 0.1 0.1 0.1

0.1 0.1

1. #

1. 4.6 3.5 2.5 2.2 2.5 3.1 4.3 3.3 4.2 5.3 7.3 7.1 8.2 13.4 15.1 7.0 1.5 4.7 4.0 4.0 2.8 2.5 2.6 3.1 3.9 5.3 6.8 9.9 10.5 9.1 6.8 7.5 6.6 4.6 1.6 N NNE NE ENE E

E S E SE SSE S

SSW SW WSW W

WNW NW NNW VAR CA L M 2.7 1.6 1.7 1.5 1.6 1.9 2.6 1.7 1.7 1.3 1.5 1.6 2.1 2.9 3.8 2.9 1.5 6.8 1.0 0.6 0.5 0.5 0.6 0.7 1.2 1.2 0.8 0.7 1.3 1.8 2.7 5.7 7.3 2.4 0.1 0.3 0.2

0.1 0.2 0.4 0.2 0.4 0.6 1.2 1.7 1.5 5.3 6.0 0.9 0.2

1. #

0.1 0.1 0.2 0.4 0.8 1.4 1.1 0.5 0.8 1.0 0.1

1. # 0.1 0.2 0.5 1.0 0.4 0.1 0.1 0.1

0.1 0.4 0.5 0.1

0.1 0.1

1. 4.2 2.4 2.2 2.0 2.3 2.9 4.3 3.4 3.6 4.4 7.0 6.7 6.9 14.8 18.2 6.3 1.6 6.8 3.5 3.6 2.6 2.4 2.9 3.6 3.6 4.7 6.7 11.1 11.8 8.6 6.0 6.9 6.7 4.3

1.7 TOTAL

39.7 30.0 18

.8 7.4 2.8 0.9 0.2 # 100.0 6.1 TOTAL 41.4 29.1 19

.0 6.7 2.5 1.1 0.2 # # 100.0 5.9

1. DENOTES LESS THAN 0.05%

1. DENOTES LESS THAN 0.05%

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-69 Table 2.3-9 Percentage Frequency D i stribution of 50-ft Wind Direction Versus Speed at HMS (1955-1970) (Continued)

COMP O S ITE OF ALL MONTHS SP EED C L ASS (MPH)

DIRECTION 0-3 4-7 8-12 13-18 19-24 25-31 32-38 39-46 46 TOTAL AVG SP EED N NNE NE ENE E E S E SE SSE S SSW SW W S W W WNW NW NNW VAR CALM 2.0 1.4 1.6 1.0 1.2 1.2 1.6 0.9 1.0 0.9 1.0 1.0 1.4 1.5 2.1 1.7 1.7 2.2 1.8 1.4 1.2 0.8 1.0 1.1 1.6 1.1 1.3 1.3 2.0 2.5 3.9 4.5 4.9 2.3 0.7 0.6 0.6 0.4 0.2 0.2 0.2 0.4 0.5 0.4 0.7 1.6 2.7 3.2 6.1 5.2 0.8 # 0.2 0.3 0.2 # # # 0.1 0.1 0.2 0.6 1.4 1.6 1.0 2.9 2.8 0.1 #

1. # # # # # # 0.1 0.3 0.8 0.5 0.2 0.9 1.3 # #

1. # #

1. # 0.2 0.3 0.2 # 0.1 0.3 #

1. 0.1 # # # # #

1. # #

1. 4.6 3.7 3.4 2.0 2.4 2.5 3.7 2.0 3.0 4.0 7.2 8.5 9.7 16.0 16.6 4.9 2.4 2.2 4.9 5.7 4.9 4.1 3.9 4.0 4.5 5.7 6.4 9.5 10.9 9.9 7.7 9.7 9.6 5.1 2.7 TOTAL 25.4 33.4 23.8 11.5 4.1 1.1 0.1 # # 100.0 7.6 # DENOTES LESS THAN 0.05%

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-70 Table 2.3-10 Percent Frequency of O cc u rrence of Wind Direction at the Hanford Reservation*

WIND DI R E CTION MONTH/YEAR/SITE/ELEVATION NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW N VARI-ABLE CALM 4/74 CGS 33' 4/74 CGS (temp) 23' 4/74 HMS 50' April (1955-1970) HMS (hist) 50' 2.64 2.64 1.53 3.2 3.06 3.61 2.36 3.2 0.42 1.11 1.25 2.0 0.83 1.81 1.53 2.3 1.53 3.75 1.94 2.0 3.19 4.72 2.64 3.1 7.50 6.81 2.50 2.5 8.89 10.42 1.81 3.2 13.06 6.11 3.61 4.7 7.78 4.31 6.81 9.5 5.69 4.72 15.28 11.2 8.75 10.28 13.19 11.8 17.08 16.11 22.08 16.6 9.31 7.50 17.08 14.3 4.17 2.36 2.78 3.7 3.19 2.92 1.53 4.2 0.97 3.19 1.67 2.1 0.00 0.00 0.42 0.4 5/74 CGS 33' 5/74 CGS (temp) 23' 5/74 HMS 50' May (1955-1970) HMS (hist) 50' 2.15 0.94 1.88 3.1 1.75 2.69 1.34 3.7 1.21 0.94 0.81 2.3 1.88 2.82 1.75 2.4 2.02 2.28 3.36 2.2 2.42 4.57 1.88 3.3 8.06 8.20 1.75 2.7 12.37 11.83 2.15 2.8 14.25 9.14 4.70 3.3 6.99 6.32 8.47 6.4 9.27 7.53 14.92 9.8 10.62 8.20 14.25 11.4 12.90 8.06 21.24 18.1 6.32 5.24 13.31 17.2 3.09 2.55 1.75 3.6 1.61 1.48 1.88 3.9 2.02 6.05 3.90 3.2 0.00 0.13 0.67 0.6 6/74 CGS 33' 6/74 CGS (temp) 23' 6/74 HMS 50' June (1955-1970) HMS (hist) 50' 2.50 3.06 1.94 3.6 2.92 4.72 2.92 3.7 4.31 3.75 2.22 2.0 3.33 4.44 2.50 2.2 3.61 7.22 3.75 2.2 5.14 7.50 4.03 2.8 7.22 7.78 2.64 2.1 7.64 6.67 2.36 2.6 7.08 6.53 3.06 3.0 4.03 5.42 4.44 6.5 3.89 3.61 7.22 8.5 7.22 4.86 6.25 11.2 9.31 10.0 24.03 19.6 7.36 10.0 19.58 18.9 3.61 4.72 4.72 3.5 2.50 2.22 2.36 3.9 0.69 7.50 5.56 3.3 0.00 0.00 0.42 0.4 7/74 CGS 33' 7/74 CGS (temp) 23' 7/74 HMS 50'

July (1955-1970) HMS (hist) 50' 2.42 3.63 2.82 3.8 3.23 4.30 1.88 3.8 2.42 3.49 2.96 2.0 4.03 3.09 2.55 2.5 3.36 5.51 3.90 2.3 5.38 5.78 3.76 2.9 7.66 8.60 2.69 1.9 9.41 9.14 3.76 2.5 7.26 10.22 2.96 2.5 6.72 3.36 5.11 6.3 4.70 3.76 12.50 8.4 6.59 4.30 12.90 9.9 8.47 5.65 16.53 19.5 9.01 12.50 11.96 18.9 5.51 5.78 2.96 3.5 4.70 3.76 2.42 4.5 4.30 6.99 8.33 4.4 0.00 0.13 0.00 0.4 8/74 CGS 33' 8/74 CGS (temp) 23' 8/74 HMS 50' August (1955-1970) HM S (hist) 50' 6.45 6.72 5.65 3.4 5.65 8.87 3.49 2.9 2.02 4.03 2.55 2.0 2.42 4.03 2.42 2.4 2.82 3.76 2.69 2.7 4.30 5.38 2.55 3.4 8.47 10.48 2.55 2.4 10.08 10.62 2.96 2.5 8.87 5.51 3.49 3.2 4.44 3.49 3.36 6.4 3.76 2.42 7.12 8.6 3.49 2.69 10.62 11.3 9.81 7.39 16.80 18.3 9.01 8.74 16.13 17.4 6.59 7.12 5.65 3.9 7.66 3.49 4.97 4.4 2.42 4.30 6.18 4.2 0.00 0.40 0.81 0.6 9/74 CGS 33' 9/73 CGS (temp) 23' 9/73 HMS 50' September (1955-1970)

HMS (hist) 50' 9.58 4.72 7.36 4.9 5.42 6.11 4.86 4.6 5.56 3.61 1.67 2.4 3.33 4.86 3.75 2.9 3.75 3.75 2.50 2.8 2.92 5.42 2.50 3.9 2.92 7.92 2.64 2.7 8.19 8.47 2.92 2.8 5.28 5.28 3.61 3.4 4.03 4.72 4.58 6.0 2.36 2.36 8.61 8.8 5.00 3.75 6.81 11.0 7.78 7.36 15.28 14.7 9.03 6.94 15.97 14.8 8.19 7.08 7.22 4.9 12.92 8.47 4.03 5.6 2.64 9.17 3.06 2.6 0.00 0.00 2.64 1.2 10/74 CGS 33' 10/73 CGS (temp) 23' 10/73 HMS 50' October (1955-1970) HMS (hist) 50' 5.65 3.76 2.42 3.5 5.78 4.70 3.76 3.6 6.59 2.69 2.15 2.3 3.23 4.30 1.75 2.8 2.28 3.63 3.23 3.3 4.84 6.85 3.63 5.1 6.05 9.68 2.96 3.4 6.85 11.96 3.23 3.5 5.65 7.53 5.65 4.5 3.49 5.51 10.89 7.4 3.76 1.75 11.16 9.2 4.97 4.84 9.27 10.7 8.60 7.12 14.78 13.2 7.12 6.85 11.42 12.9 8.06 4.97 4.84 5.4 8.60 3.23 2.82 4.7 2.69 8.20 1.48 1.8 0.13 0.27 4.57 2.7

• For some months, when concurrent measurements are not available for all si tes shown; previous y ear data is given.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-71 Table 2.3-10 Percent Frequency of O cc u rrence of Wind Direction at the Hanford Res e rvation* (Continued)

WIND D I RECTION MONTH/YEAR/

S I T E/ELEVATION NNE NE ENE E E S E SE SSE S S S W SW WSW W WNW NW NNW N VARI-ABLE CALM 11/74 C G S 33' 11/73 C G S (t e m p) 2 3' 11/73 HMS 50' Nove mber (1955-1970) HMS (hist) 50' 5.28 2.22 3.33 3.4 3.61 4.86 1.67 2.6 4.31 1.81 1.81 2.3 1.11 1.53 1.81 2.5 2.08 4.44 3.61 3.1 4.72 8.19 6.11 4.4 7.36 9.72 3.06 3.3 10.42 8.19 5.00 4.2 10.69 5.42 6.25 5.2 6.94 2.22 7.50 7.4 3.33 2.22 4.86 7.1 4.86 5.69 5.42 8.4 7.08 16.25 13.89 13.4 10.42 8.06 19.31 15.1 8.33 3.75 5.97 7.0 6.94 2.92 4.44 4.6 1.81 3.61 0.56 1.4 0.00 0.28 5.42 4.6 12/74 C G S 33' 12/73 C G S (t e m p) 2 3' 12/73 HMS 50' Dece mber (1955-1970) HMS (hist) 50' 2.28 2.28 2.02 2.5 2.15 5.38 1.48 2.3 1.21 2.28 2.15 1.9 0.94 2.82 2.28 2.3 1.34 3.76 2.28 2.8 2.82 7.53 3.49 4.2 8.06 9.95 3.49 3.4 13.44 10.62 4.30 3.5 11.02 5.24 4.70 4.4 4.84 3.09 7.12 7.1 6.05 2.02 8.20 6.8 6.99 4.30 9.54 6.9 10.89 9.41 13.84 14.8 12.23 10.48 20.02 18.2 8.33 4.84 4.57 6.3 4.57 2.82 1.75 4.2 1.61 7.93 2.15 1.6 0.00 0.67 6.59 6.8 1/75 C G S 33' 1/74 C G S (te m p) 23' 1/74 HMS 50' January (1955-1970)

HMS (hist) 50' 4.57 2.28 2.69 3.4 3.76 1.88 3.09 2.9 4.44 1.08 2.02 1.9 2.42 1.88 1.34 2.4 3.09 3.09 2.55 2.8 3.63 4.97 4.57 4.7 6.45 7.12 4.03 3.1 6.18 13.17 4.03 3.6 6.99 17.34 5.24 4.9 6.45 8.20 16.67 6.7 6.05 3.23 13.04 6.0 3.76 3.23 7.66 6.1 6.85 5.51 7.53 14.2 11.96 6.32 9.81 19.5 13.71 3.49 4.30 6.4 5.91 1.88 3.09 4.6 1.48 4.70 1.21 1.4 0.00 1.61 4.70 5.4 2/75 C G S 33' 2/74 C G S (te m p) 23' 2/74 HMS 50' February (1955-1970)

HMS (hist) 50' 4.76 1.04 2.08 4.4 2.38 1.79 2.68 3.5 1.93 1.34 2.08 1.9 0.89 1.64 1.19 2.1 1.64 4.32 4.17 2.2 2.98 7.44 4.46 3.4 7.74 9.08 2.83 2.7 7.89 13.54 5.36 2.9 8.18 13.39 8.63 4.2 4.91 6.25 10.57 7.5 4.61 7.14 12.95 7.4 3.57 5.21 10.86 9.3 5.21 8.33 10.12 15.4 18.76 5.65 10.71 18.4 14.43 2.38 4.91 5.8 8.18 1.34 2.38 5.0 1.19 8.93 2.23 1.5 0.00 0.19 1.79 2.4 3/75 C G S 33' 3/74 C G S (te m p) 23' 3/74 HMS 50' March (1955-1970)

H M S (hist) 50' 2.82 1.61 1.75 4.5 1.34 2.55 2.69 3.5 1.21 2.02 2.02 1.7 1.34 2.82 1.61 2.0 2.02 1.61 1.21 2.3 2.42 6.59 4.84 3.7 7.26 6.72 3.76 2.8 10.62 10.08 4.30 3.8 8.33 10.08 7.12 5.4 9.81 7.66 15.05 9.9 4.97 4.84 10.75 10.4 2.15 5.65 9.01 9.6 6.99 6.72 11.16 14.4 11.16 6.72 13.04 13.7 11.56 4.57 5.65 5.0 5.91 3.23 2.82 5.1 1.61 9.41 2.82 1.5 0.00 1.08 0.40

0.7 April

1974-March 1 9 75 C G S 33' 1955-1970 HMS (hi s t) 50' 4.25 3.7 3.42 3.4 2.97 2.0 2.16 2.4 2.47 2.6 3.73 3.7 7.07 2.8 9.35 3.2 8.89 4.1 5.88 7.2 4.89 8.5 5.67 9.8 9.27 16.0 10.08 16.6 7.93 4.9 6.04 4.5 1.96 2.4 0.01 2.2

• For some months, when concu rrent measurements are n o t availab l e for all s i tes shown; pre v ious y ear data is given.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-72 Table 2.3-11 Persistence of Wind Di r ection in One Sector (22.5 Degrees) from 4/74 through 3/75 at 33-ft Level (Stabil i ty Based On Te m p erature Di f f erence) DATE STA R TED DAY HOUR WIND DIR HOU R S OF PE R S I S TEN C E HOURS EACH S T ABI L ITY AVERAGE S P EED (MPH) 400 22 NW 14 0 0 3 11 0 V UNS UN S TA NEUTR M S T A V S T A .00

.00 10.30 11.08 .00 171 10 S

10 0 1 7

2 0 UNKNO

V UNS UN S TA NEUTR M S T A .00 16.00 18.14 15.00 .00

295 15

NNW

10 0 0 0 1

2 V S T A UNKNO

V UNS UN S TA NEUTR .00

.00 .00 19.57 21.18

327 14

NW

10 7 0

0 0 0 M S T A V S T A UNKNO

V UNS UN S TA 9.98 .00

.00 .00

.00

425 6

NNW

10 5 5

0 0 0 NEUTR M S T A V S T A UNKNO

V UNS 3.81 4.06 .00

.00 .00 1 7 2

0 0 UN S TA NEUTR M S T A V S T A UNKNO 5.53 5.84 6.81 .00

.00 134 7 SS W 9 0 1 8

0 0 0 V UNS UN S TA NEUTR M S T A V S T A UNKNO .00 23.78 19.54 .00 .00 .00 219 17 WNW 9 0 2

2 5

0 V UNS UN S TA NEUTR M S T A V S T A .00 17.34 22.97 16.69 .00 393 2 NW 9 0 0

9 0

V UNS UN S TA NEUTR M S T A .00

.00 #999.0 (m issing data)

.00

404 2

NNW

9 0 0 0 0

1 V S T A UNKNO

V UNS UN S TA NEUTR .00

.00 .00

.00 13.63 8 0 0 M S T A V S T A UNKNO 11.18 .00

.00 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-73 Table 2.3-11 Persistence of Wind Dir ection in One Sector (22.5 Degrees) from 4/74 through 3/75 at 33-ft Level (Continued) (Stabil i ty Based on Temperature Di f f erence) DATE STA R TED DAY HOUR WIND DIR HOU R S OF PE R S I S TEN C E HOURS EACH S T ABI L ITY AVERAGE S P EED (MPH) 407 5 SS W 9 0 0 V UNS UN S TA .00

.00

102 13

WNW

8 2 7

0 0 0 NEUTR M S T A V S T A UNKNO

V UNS 17.20 20.64 .00

.00 .00 5 1 2

0 0 UN S TA NEUTR M S T A V S T A UNKNO 16.67 19.07 16.91 .00

.00 132 13 WNW 8 0 6 1

1 0

0 V UNS UN S TA NEUTR M S T A V S T A UNKNO .00 22.17 22.06 17.06 .00

.00 244 11 NNE 8 0 7

1 0

0 V UNS UN S TA NEUTR M S T A V S T A .00 15.35 12.87 .00

.00 271 16 NW 8 0 0 1

2 5 UNKNO

V UNS UN S TA NEUTR M S T A .00 .00 18.52 16.92 11.63

363 10

S

8 0 0 0 0

5 V S T A UNKNO

V UNS UN S TA NEUTR .00

.00 .00

.00 13.70

396 21

NW

8 3 0

0 0 0 M S T A V S T A UNKNO

V UNS UN S TA 13.80 .00

.00 .00

.00

401 12

NNW

8 4 4

0 0

0 0 NEUTR M S T A V S T A UNKNO V UNS UN S TA 9.10 8.16 .00

.00

.00

.00

402 6

N

8 5 3 0 0 0 NEUTR M S T A V S T A UNKNO

V UNS 12.80 11.49 .00

.00 .00 0 7 1

0 0 UN S TA NEUTR M S T A V S T A UNKNO .00 9.04 9.99 .00

.00 426 14 S W 8 0 2 3

3 0

0 V UNS UN S TA NEUTR V S T A V S T A UNKNO .00 18.58 16.88 15.28 .00

.00 430 8 NNW 8 0 6

1 1

0 V UNS UN S TA NEUTR M S T A V S T A .00 10.54 10.68 7.82 .00 0 UNKNO .00 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-74 Table 2.3-12 Persistence of Wind Direction in Two Sectors (45 Degrees) from 4.74 through 3/75 at CGS for 33-ft Level (Stability Based on Temperature Difference)

DATE STARTED DAY HOUR WIND DIRECTIONS HOURS OF PERSISTENCE HOURS EACH STABILITY AVERAGE SPEED (MPH) 400 22 NW NNW 26 0 0 8 15 3 V UNS UNSTA NEUTR M STA V STA .00 .00 11.80 11.05 6.23 379 12 NW NNW 23 0 0 0 6 17 UNKNO V UNS UNSTA NEUTR M STA .00 .00

.00 3.73 3.78 399 23 NW NNW 22 0 0 0 0 7 V STA UNKNO V UNS UNSTA NEUTR .00

.00 .00 .00 15.79 424 21 NW NNW 20 15 0 0 0 2 M STA V STA UNKNO V UNS UNSTA 12.65 .00 .00 .00 4.20 102 1 NNW NW 19 8 10 0 0 0 NEUTR M STA V STA UNKNO V UNS 5.88 7.18 .00

.00 .00 8 2 9 0 0 UNSTA NEUTR M STA V STA UNKNO 18.20 28.03 21.20 .00 .00 400 17 WNW NW 19 0 0 3 16 0 0 V UNS UNSTA NEUTR M STA V STA UNKNO .00 .00 10.30 11.44 .00 .00 244 5 N NNE 18 0 7 7 4 0 V UNS UNSTA NEUTR M STA V STA .00 15.35 15.01 9.66 .00 404 2 NNW N 18 0 0 0 9 9 UNKNO V UNS UNSTA NEUTR M STA .00 .00

.00 11.56 10.83 431 10 NW NNW 18 0 0 0 5 5 V STA UNKNO V UNS UNSTA NEUTR .00

.00 .00 17.79 17.77 171 3 S SSW 17 8 0 0 1 8 M STA V STA UNKNO V UNS UNSTA 8.45 .00 .00 16.00 18.37 224 17 NNW NW 17 5 3 0 0 0 NEUTR M STA V STA UNKNO V UNS 16.60 13.33 .00

.00 .00 4 3 10 0 0 UNSTA NEUTR M STA V STA UNKNO 10.07 14.39 12.00 .00

.00 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-75 Table 2.3-12 Persistence of Wind Dir ection in Two Sectors (45 Degrees) from 4.74 through 3/75 at CGS for 33-ft Level (Continued) (Stabil i ty Based on Temperature Di f f erence) DATE STA R TED DAY HOUR WIND DI R E CTIONS HOU R S OF PE R S I S TEN C E HOURS EACH S T ABI L ITY AVERAGE S P EED (MPH) 401 21 NNW N 17 0 0 7 3

7 0 V UNS UN S TA NEUTR M S T A V S T A UNKNO .00

.00 9.04 8.95 6.78 .00 406 21 SSW SW 17 0 0

2 15 0 V UNS UN S TA NEUTR M S T A V S T A .00

.00 17.20 21.52 .00 438 1 SSE S 17 0 0 2

2 7 UNKNO

V UNS UN S TA NEUTR M S T A .00 .00 7.67 5.47 7.21

396 17

WNW NW

16 6 0 0 0 10 V S T A UNKNO

V UNS UN S TA NEUTR 8.80 .00 .00

.00 8.80

423 8

NW NNW

16 6 0

0 0 0 M S T A V S T A UNKNO

V UNS UN S TA 7.94 .00

.00 .00

.00

98 11

WNW NW

15 10 6 0

0 0 3 NEUTR M S T A V S T A UNKNO

V UNS UN S TA 7.61 5.58 .00

.00 .00 15.67

253 15

WNW NW

15 4 8

0 0 0 NEUTR M S T A V S T A UNKNO

V UNS 14.00 11.62 .00

.00 .00 0 2 9

4 0 UN S TA NEUTR M S T A V S T A UNKNO .00 16.41 11.59 7.27 .00 293 11 W WNW 15 0 4 3

8 0

0 V UNS UN S TA NEUTR M S T A V S T A UNKNO .00 16.25 18.27 9.39 .00

.00 355 6 WSW W 15 0 0

7 8

0 V UNS UN S TA NEUTR M S T A V S T A .00

.00 25.65 15.14 .00 0 UNKNO .00 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-76 Table 2.3-12a Longest Persistence of Wind Direction in On e (22.5 Degrees) and Two (45 Degrees) Sectors During First and Second A nnual Cycles at 33-ft Level (Stability Based on Temperature Difference)

First Annual Cycle (April '74 - March '75)

Second Annual Cycle (April '75 - March '76)

MONTH WIND DIRECTION HOURS OF PERSISTENCE HOURS OF EACH STABILITY AVERAGE WIND SPEED MONTH WIND DIRECTION HOURS OF PERSISTENCE HOURS OF EACH STABILITY AVERAGE WIND SPEED January NW 14 0 0

3 11 0 0

V UNS UNSTA NEUTR M STA V STA UNKNO

.00

.00 10.30 11.08 .00 .00 February NNE 33 0 3 10 20 0 0 V UNS UNSTA NEUTR M STA V STA UNKNO

.00 30.22 29.04 23.15 .00 .00 January NW, NNW 26 0 0

8 15 3 0

V UNS UNSTA NEUTR M STA V STA UNKNO

.00

.00 11.86 11.05 6.23 .00 January N, NNE 35 0 3 10 22 0 0

V UNS UNSTA NEUTR M STA V STA UNKNO

.00 30.22 29.04 21.84 .00

.00 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-77 Table 2.3-13 Percent Frequency of Occurrence of Wind Speed at the Hanford Reservation (1)

Wind Speed Range, mph (2) Month/Year/Site/Elevations

Calm 1-3 4-7 8-12 13-18 19-24 25-Up Average Speed (mph) 4/74 CGS 33'

4/74 CGS (temp) 23'

4/74 HMS 50'

April (1955-1970) HMS (hist) 50' 0.00 0.00 0.42 9.58 13.19 9.72 16.8 31.11 37.50 26.11 31.6 24.17 23.47 33.33 26.6 18.19 17.78 19.72 16.2 6.25 4.72 7.64 6.6 1.81 0.42 3.06 2.2 9.8 8.7 10.3 9.0 5/74 CGS 33'

5/74 CGS (temp) 23' 5/74 HMS 50' May (1955-1970) HMS (hist) 50' 0.00 0.13 0.67 12.50 14.11 12.63 16.6 36.16 38.17 30.11 33.3 33.33 28.23 32.66 27.9 12.77 11.96 18.15 15.1 3.63 2.55 5.38 6.0 0.54 0.67 0.40 1.1 8.4 7.9 9.0 8.8 6/74 CGS 33'

6/74 CGS (temp) 23'

6/74 HMS 50'

June (1955-1970) HMS (hist) 50' 0.00 0.00 0.42 11.25 21.67 13.61 11.7 44.03 47.64 35.42 34.6 23.89 18.06 26.11 28.1 9.72 9.03 15.69 17.4 3.61 3.06 7.64 6.8 1.11 0.56 1.11 1.4 8.5 6.9 9.0 9.2 7/74 CGS 33'

7/74 CGS (temp) 23'

7/74 HMS 50' July (1955-1970) HMS (hist) 50' 0.00 0.13 0.00 16.13 25.40 16.26 15.0 43.95 45.30 38.44 37.8 24.33 19.89 27.28 26.2 8.74 8.06 13.44 14.2 1.61 1.21 4.57 5.8 0.40 0.00 0.00 1.0 7.2 6.4 8.1 8.6 8/74 CGS 33'

8/74 CGS (temp) 23'

8/74 HMS 50' August (1955-1970) HMS (hist) 50' 0.00 0.40 0.81 21.91 27.82 16.53 18.9 46.37 46.24 43.28 38.2 19.35 17.20 26.21 25.6 7.12 6.85 7.53 12.3 3.36 1.21 5.11 4.2 0.13 0.00 0.54 0.8 6.8 6.0 7.5 8.0 9/74 CGS 33'

9/73 CGS (temp) 23'

9/73 HMS 50' September (1955-1970) HMS (hist) 50' 0.00 0.00 2.64 27.78 26.67 20.42 23.6 38.06 42.08 36.25 36.8 22.50 22.50 28.33 23.8 8.75 7.50 8.75 10.9 1.39 1.25 3.19 4.1 0.42 0.00 0.42 0.8 6.5 6.2 7.1 7.5 (1) For some months, when, concurrent measurements are not available for all sites shown; previ ous year data is given.

(2) HMS (hist) 50' calm values are incl uded in the 1-3 mph range group.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-78 Table 2.3-13 Percent Frequency of Occurrence of Wind Speed at the Hanford Reservation (1) (Continued)

Month/Year/Site/Elevation

Calm 1-3 4-7 8-12 13-18 19-24 25-Up Average Speed (mph) 10/74 CGS 33'

10/73 CGS (temp) 23'

10/73 HMS 50' October (1955-1970) RMS (hist) 50' 0.13 0.27 4.57 44.49 30.91 25.81 32.1 34.27 40.99 34.95 34.0 11.16 16.80 23.39 21.3 3.36 8.87 7.53 8.3 0.94 2.15 3.23 3.3 0.00 0.00 0.54 1.0 4.8 6.2 6.7 6.7 11/74 CGS 33'

11/73 CGS (temp) 23'

11/73 HMS 50' November (1955-1970) HMS (hist) 50' 0.00 0.28 5.42 34.72 17.92 19.86 39.7 39.58 37.08 33.89 30.1 16.11 21.53 26.67 18.9 5.97 7.78 6.81 7.5 1.53 2.36 5.28 2.7 0.00 0.00 2.08 1.1 5.8 7.1 7.5 6.2 12/74 CGS 33' 12/73 CGS (temp) 23'

12/73 HMS 50' December (1955-1970) HMS (hist) 50' 0.00 0.67 6.59 31.05 41.53 25.13 41.4 37.23 31.72 31.05 29.2 21.51 15.86 23.79 18.9 6.05 8.74 9.95 6.7 1.61 1.21 3.09 2.5 1.21 0.27 0.40 1.3 6.4 5.7 6.7 6.0 1/75 CGS 33'

1/74 CGS (temp) 23'

1/74 HMS 50' January (1955-1970) HMS (hist) 50' 0.00 2.02 4.70 29.44 29.97 21.10 38.8 36.83 24.46 26.48 30.2 19.76 15.46 21.24 19.5 5.38 16.40 13.31 7.6 2.02 7.80 7.66 2.7 0.67 3.90 5.51 1.2 6.4 8.7 9.3 6.4 2/75 CGS 33'

2/74 CGS (temp) 23'

2/74 HMS 50' February (1955-1970) HMS (hist) 50' 0.00 1.19 1.79 22.47 26.64 20.09 30.8 36.90 29.17 33.63 31.0 23.36 22.17 23.66 23.9 9.82 15.03 15.63 9.3 5.21 4.46 4.61 3.2 1.49 1.34 0.60 1.7 7.8 7.6 8.0 7.0 3/75 CGS 33'

3/74 CGS (temp) 23'

3/74 HMS 50'

March (1955-1970) HMS (hist) 50' 0.00 1.08 0.40 13.04 28.90 16.94 20.0 31.85 31.32 33.33 32.6 27.02 18.68 25.81 26.1 14.11 12.77 13.04 13.5 4.44 3.36 7.66 5.4 1.08 3.36 2.82 2.4 8.7 7.8 9.1

8.4 April

1974 - March 1975 CGS 33'

1955-1970 HMS (hist) 50' 0.01 22.89 25.4 38.04 33.3 22.20 23.9 9.14 11.6 2.95 4.4 0.73 1.4 7.2 7.6 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-79 Table 2.3-14 Diurnal Variation of 33-ft Ele v ation Dry Bulb T e mperature ( F) at CGS and Mon t hly Average Dry Bulb Temperature

( F) at the Hanford R e servation Mont h/Year Hour April 19 7 4 May 19 7 4 June 19 7 4 Ju ly 19 7 4 Aug u st 19 7 4 September 19 7 4 October 19 7 4 Nove m b er 19 7 4 December 19 7 4 Janu a ry 19 7 5 Feb r uary 19 7 5 March 19 7 5 Annual A v erage (April 1 974 -March 19 7 5) 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 46.7 46.1 44.9 44.4 43.3 43.6 45.3 49.2 51.8 54.2 56.2 57.9 59.3 60.5 60.9 60.9 60.3 58.8 56.1 53.6 51.8 50.3 48.9 47.7 51.0 49.8 48.3 46.9 45.8 47.0 50.9 54.4 56.8 59.0 61.3 63.1 64.9 66.3 67.5 67.7 67.4 66.2 63.9 60.3 57.7 56.0 54.2 52.7 63.7 62.1 60.3 58.1 57.5 59.6 63.8 68.0 72.7 75.3 78.0 80.5 82.3 83.8 85.1 84.6 85.1 84.1 81.9 77.8 73.7 70.7 68.0 66.0 66.5 64.5 62.2 60.8 59.7 61.0 65.3 69.3 73.0 75.1 77.3 79.4 81.3 83.2 84.7 85.2 85.2 84.7 82.7 78.7 75.5 73.0 70.8 68.9 66.8 64.6 63.0 61.2 60.0 59.6 63.6 68.8 72.9 76.4 79.1 81.6 84.0 86.2 87.5 88.4 88.4 87.1 83.7 79.6 75.7 73.6 71.0 69.0 58.2 57.1 56.3 55.3 54.1 53.0 53.9 58.9 64.8 68.6 72.0 75.2 77.8 79.8 81.0 81.6 81.2 78.4 73.6 69.4 67.1 64.5 62.3 60.4 45.6 44.7 43.3 42.6 41.3 40.8 40.4 43.4 48.3 52.7 56.4 59.5 62.0 63.7 64.8 65.1 63.7 60.0 56.4 53.5 51.3 49.4 48.2 46.4 40.3 40.1 39.6 39.5 39.0 38.7 38.3 38.3 39.9 41.6 43.8 45.4 46.7 47.4 47.6 46.9 45.0 43.5 42.7 42.1 41.6 41.0 40.2 40.0 35.3 35.3 35.1 35.0 34.9 34.1 33.8 33.8 34.4 35.8 37.9 39.6 41.0 41.7 41.5 40.5 38.8 37.7 36.8 36.8 36.8 36.1 35.2 35.1 30.9 30.8 30.4 30.6 30.1 30.0 29.4 29.5 30.3 31.7 33.0 34.9 35.7 36.5 37.2 36.7 35.2 33.9 32.9 32.1 31.5 31.1 31.0 30.8 30.6 30.5 30.1 30.2 30.1 29.8 29.6 29.3 31.3 33.9 35.6 37.6 38.6 39.3 39.7 39.5 38.7 36.8 35.3 34.4 33.8 32.8 32.4 31.6 37.5 36.6 35.9 35.0 34.9 34.7 35.1 37.4 40.6 42.9 45.2 46.9 48.6 49.4 50.3 50.8 50.0 47.9 44.9 42.7 41.2 40.1 39.2 38.1 47.9 46.9 45.9 45.1 44.3 44.4 45.8 48.4 51.2 54.3 56.2 58.4 60.1 61.6 62.4 62.3 61.7 60.4 57.7 55.2 53.3 51.7 50.2 49.0 Month l y Av e rage Dry Bulb Tempera t ure ( F)* (Site, Elevation)

CGS 33' CGS 7' CGS (temp) 3' HMS 3' 1950-1 970 HMS (hist) 3' 52.2 52.7 53.3 52.5 52.5 57.4 58.3 59.6 57.9 61.8 72.5 73.0 74.2 73.3 69.9 73.6 74.3 75.3 74.8 77.5 74.7 75.0 76.3 76.3 75.3 66.9 66.3 (65.0) 68.3 67.0 51.7 50.6 (51.2) 52.0 53.2 42.1 41.9 (39.7) 42.1 40.1 36.8 36.1 (37.8) 35.7 33.4 32.3 32.2 (29.0) 32.0 30.3 33.8 33.9 (40.7) 33.6 37.5 41.9 42.2 (45.4) 42.0 44.0 53.1 53.1 Not Computed 53.4 *For some months, when concurrent measurements are not available for all sites show n, former year data is given in parentheses.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-80 Table 2.3-15 Diurnal Variation of 33-ft Ele v ation Wet Bulb T e mperature

( F) at CGS and Mon t hly Average Wet Bulb T e mperature

( F) at the Hanford R e servation Mont h/Year Hour April 19 7 4 May 19 7 4 June 19 7 4 Ju ly 19 7 4 Aug u st 19 7 4 September 19 7 4 October 19 7 4 Nove m b er 19 7 4 December 19 7 4 Janu a ry 19 7 5 Feb r uary 19 7 5 March 19 7 5 Annual A v erage (April 1 974 -March 19 7 5) 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 41.9 41.5 40.9 40.7 40.2 40.5 41.9 44.2 45.5 46.6 47.1 47.8 48.2 48.6 48.7 48.6 48.5 48.0 46.8 45.8 44.8 43.8 43.1 42.5 44.4 43.8 43.3 42.6 41.9 42.7 44.9 46.1 47.5 48.1 49.1 49.7 50.3 50.7 51.0 50.9 50.8 50.4 49.7 48.5 47.4 46.6 45.8 45.1 52.1 51.6 50.9 50.1 50.0 51.2 53.3 54.9 56.8 57.7 58.8 59.6 60.0 60.3 60.7 60.1 60.5 60.3 59.6 58.4 56.5 54.9 54.1 53.1 54.7 54.1 53.1 52.6 52.3 53.4 55.4 56.9 57.9 58.5 59.0 59.7 60.2 60.4 60.9 61.1 61.0 60.8 60.1 59.0 58.0 57.2 56.2 55.6 54.9 54.1 53.5 52.8 52.5 52.3 54.5 56.8 58.2 59.7 60.6 61.2 61.7 62.2 62.4 62.6 62.6 62.0 60.9 59.6 58.0 57.3 56.6 55.8 49.1 48.6 48.0 47.6 47.0 46.4 47.1 49.7 52.6 54.2 55..4 56.5 57.4 57.9 58.2 58.3 58.0 57.0 55.4 53.6 52.4 51.4 50.5 49.7 40.7 40.1 39.3 38.9 38.0 37.8 37.5 39.5 42.7 44.9 46.7 48.0 49.1 49.7 50.2 50.3 49.8 48.4 46.4 44.8 43.4 42.4 41.8 41.1 38.0 38.0 37.6 37.6 37.3 37.1 36.8 36.9 38.2 39.3 40.8 41.6 42.3 42.6 42.8 42.4 41.4 40.5 40.0 39.6 39.0 38.5 37.8 37.7 33.5 33.6 33.5 33.4 33.2 32.7 32.4 32.2 32.6 34.0 35.4 36.4 37.3 37.6 37.6 36.9 35.8 35.1 34.4 34.4 34.4 34.1 33.5 33.5 29.0 29.0 28.7 28.9 28.5 28.4 27.9 27.8 28.4 29.6 30.6 31.8 32.3 32.8 33.2 32.9 31.9 30.9 30.4 29.7 29.4 29.0 29.0 28.8 28.9 28.7 28.5 28.4 28.3 28.0 27.8 27.6 29.5 31.2 32.1 33.2 33.9 34.2 34.5 34.4 34.0 32.9 32.0 31.4 31.1 30.5 30.3 29.6 34.0 33.4 32.9 32.3 32.2 32.1 32.4 34.3 36.1 37.1 38.3 39.0 39.5 40.0 40.4 40.6 40.1 39.2 37.8 36.8 35.9 35.2 34.6 34.0 41.8 41.4 40.9 40.6 40.2 40.3 41.0 42.3 43.8 45.3 46.1 47.0 47.7 48.2 48.4 48.3 48.0 47.2 46.2 45.2 44.3 43.5 42.8 42.3 Month l y Av e r age Wet Bulb Te m p erature ( F)* (Site, El e v ation)

CGS 33' CGS (temp) 3'

HMS 3' 1950-1970 HMS (hist) 3' 44.7 45.9 43.9 42.8 47.2 49.8 46.5 49.1 56.0 60.0 54.5 54.5 57.4 61.0 56.3 57.9 58.0 62.6 57.0 57.3 52.6 (54.6) 52.0 52.6 43.8 (45.5) 42.0 45.4 39.3 (37.8) 38.0 36.4 34.5 (36.4) 33.0 31.2 30.0 (26.4) 30.0 27.9 30.9 (36.6) 31.0 33.6 36.2 (39.3) 36.0 37.3 44.3 Not Calculated 43.4 *For some months, when concurrent measurements are not available for all sites show n, former year data is given in parentheses.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-81 Table 2.3-16 Diurnal Variation of 33 ft Ele v ation Dew Point Temperatu r e ( F) at C G S and Monthly Average Dew Point Temperatu r e ( F) at the Hanford Reservation Mont h/Year Hour April 19 7 4 May 19 7 4 June 19 7 4 Ju ly 19 7 4 Aug u st 19 7 4 September 19 7 4 October 19 7 4 Nove m b er 19 7 4 December 19 7 4 Janu a ry 19 7 5 Feb r uary 19 7 5 March 19 7 5 Annual A v erage (April 1 974 -March 19 7 5) 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 36.5 36.5 36.6 36.8 36.7 37.0 38.3 38.8 38.8 38.8 37.8 37.3 36.8 36.4 35.9 35.8 36.0 36.6 37.0 37.3 37.1 36.8 36.7 36.6 37.3 37.5 37.8 37.9 37.7 38.1 38.6 37.3 37.8 37.2 36.7 36.2 35.5 35.1 34.6 34.0 34.1 34.5 35.2 36.3 36.6 36.8 37.1 37.1 42.0 42.5 42.7 43.1 43.3 44.1 44.6 44.3 44.4 44.2 44.2 43.8 43.3 42.6 42.3 41.5 41.9 42.4 42.8 43.5 42.7 41.7 42.4 42.1 45.1 45.4 45.5 45.7 46.2 47.2 47.5 47.5 46.7 46.0 45.7 45.3 44.4 43.3 43.1 43.1 42.7 42.9 42.6 43.5 44.4 44.7 44.5 44.9 45.4 45.6 45.8 45.9 46.3 46.4 47.3 47.8 47.3 47.7 47.4 46.7 46.0 45.3 44.5 44.0 44.0 43.7 44.1 44.5 44.4 44.6 45.2 45.3 40.5 40.6 39.9 40.1 40.1 39.8 40.4 41.0 41.8 41.7 41.4 40.9 40.4 39.7 39.4 39.0 38.5 38.6 39.4 39.2 38.8 38.9 39.2 39.5 35.0 34.6 34.4 34.6 34.0 34.2 34.0 35.0 36.3 36.3 36.2 36.0 35.6 35.3 35.1 35.1 35.5 36.1 35.3 34.8 34.0 33.8 34.2 34.7 35.4 35.5 35.4 35.6 35.4 35.3 35.1 35.2 36.4 36.7 37.4 37.5 37.4 37.3 37.7 37.4 37.4 37.1 37.0 36.7 36.2 35.5 35.1 35.0 31.2 31.4 31.4 31.4 31.0 30.8 30.5 30.0 30.3 31.7 32.2 32.3 32.5 32.1 32.4 32.1 31.8 31.5 31.1 31.0 31.3 31.2 31.1 31.2 26.0 26.1 25.9 26.0 26.0 25.8 25.5 25.1 25.4 26.4 26.9 27.0 27.2 27.5 27.4 27.4 27.0 26.4 25.9 26.0 25.8 25.6 25.6 26.1 26.0 25.9 25.4 25.4 25.2 24.9 24.9 26.8 27.1 26.9 26.8 27.0 27.0 27.0 27.2 27.2 27.1 27.2 27.1 27.2 27.1 27.2 26.6 28.6 28.6 28.3 28.4 28.4 28.2 28.5 30.0 29.6 28.8 28.3 27.6 27.2 26.6 26.3 26.3 25.9 26.7 27.6 28.3 28.3 27.9 27.9 27.9 35.8 35.9 35.9 36.0 35.9 36.1 36.3 36.5 36.8 37.1 36.8 36.5 36.1 35.7 35.5 35.2 35.2 35.4 35.5 35.7 35.6 35.4 35.6 35.6 Month l y Av e r age Dew Point Tempera t ure ( F) (Site/Elevation)

CGS 33' HMS 3' 19 5 0-1 970 H MS (hi s t) 3' 36.6 33.3 30.4 36.6 34.0 36.0 43.0 38.2 41.2 44.9 41.0 42.3 45.6 43.2 42.8 39.9 38.9 39.5 35.0 31.0 36.9 36.3 33.9 31.1 31.4 29.2 27.5 26.3 26.0 23.2 26.5 25.5 27.4 27.9 26.0 27.3 35.9 33.4 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000 LDC N-9 9-0 0 0 2.3-82 Table 2.3-17 Frequency of Occurrence, Dry Bulb Temperatu r e ( F) Versus Time of D a y from 4/74 through 3/75 for 33-ft Level

TI M E O F D A Y -20

-15 -20

-15

-10 1 0

0 0 2 0

0 0 3 0

0 0 4 0

0 0 5 0

0 0 6 0

0 0 7 0

0 0 8 0

0 0 9 0

0 0 10 0 0

0 11 0 0

0 12 0 0

0 13 0 0

0 14 0 0

0 15 0 0

0 16 0 0

0 17 0 0

0 18 0 0

0 19 0 0

0 20 0 0

0 21 0 0

0 22 0 0

0 23 0 0

0 24 0 0

0 T O T A L 0 0

0 5 0 5 10 15 -5 0 5 10 15 20 0 0

0 0

1 2 0 0

0 0

1 2 0 0

0 0

0 3 0 0

0 0

0 3 0 0

0 0 0 7 0 0 0 0 0

5 0 0

0 0

0 7 0 0

0 0

0 8 0 0

0 0

0 4 0 0

0 0

0 3 0 0

0 0

0 3 0 0

0 0

0 2 0 0

0 0

0 1 0 0

0 0

0 1 0 0

0 0

0 1 0 0

0 0

0 1 0 0

0 0

0 1 0 0

0 0

0 1 0 0

0 0

0 1 0 0

0 0

0 2 0 0

0 0

0 3 0 0

0 0

1 3 0 0

0 0

1 3 0 0

0 0

1 2 0 0

0 0

5 69 20 25 30 35 40 45 25 30 35 40 45 50 11 22 44 42 40 51 11 26 46 39 45 52 13 27 43 45 48 48 13 29 44 46 54 46 14 30 42 48 57 48 13 37 41 44 57 43 17 28 41 46 45 42 14 27 31 39 36 38 10 25 26 24 35 38 7 17 26 27 28 37 5 13 25 31 28 35 4 9 21 30 31 32 4 9 12 32 28 34 3 9 14 21 37 29 3 9 11 23 36 29 3 8 14 26 34 29 4 9 20 28 26 41 6 11 19 32 34 36 7 10 29 37 35 33 9 8 37 35 35 29 9 9 41 34 38 34 9 11 40 45 33 39 9 13 47 40 43 37 10 20 45 38 46 39 208 416 759 852 929 919 50 55 60 65 70 75 55 60 65 70 75 80 33 31 27 25 26 7 30 33 29 26 20 2 34 32 32 25 11 1 40 31 32 21 3 0 31 33 29 20 2 0 38 28 30 23 2 0 28 34 28 25 18 2 28 30 28 28 22 17 34 17 24 30 18 16 35 24 20 31 33 18 33 33 20 25 31 29 32 34 27 19 26 37 32 33 22 27 27 20 33 30 22 32 24 22 36 25 24 26 26 26 39 21 24 24 27 24 25 24 24 24 25 21 23 30 23 17 29 19 26 31 21 25 16 31 41 27 19 27 30 17 40 27 21 29 23 27 34 26 33 23 26 23 32 29 28 24 30 20 35 29 26 35 20 12 792 692 613 611 515 391 80 85 90 95 100 105 85 90 95 100 105 110 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 1 0

0 0

0 0 16 1 0

0 0

0 15 10 1 0

0 0 15 16 5 0

0 0 18 19 9 1

0 0 37 12 20 3 0

0 34 17 20 9 0

0 20 31 16 14 2 0 17 30 21 14 3 0 22 26 20 16 3 0 27 19 18 13 2 0 16 23 13 7 1

0 22 16 8 0

0 0 19 8 0

0 0

0 14 2 0

0 0

0 5 1

0 0

0 0 4 0

0 0

0 0 302 231 151 77 11 0 110 UN K N O T O T A L 0 3 365 0 3 365 0 3 365 0 3 365 0 4 365 0 4 365 0 4 365 0 18 365 0 47 365 0 33 365 0 18 365 0 14 365 0 12 365 0 8 365 0 7 365 0 6 365 0 6 365 0 6 365 0 3 365 0 3 365 0 3 365 0 3 365 0 3 365 0 3 365 0 217 87 6 0 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000LDCN-99-000 2.3-83 Table 2.3-18 Frequency of Occurrence, Wet Bulb Temperature (F) Versus Time of Day from 4/74 through 3/75 for 33-ft Level

TIME OF DAY

-20

-15 -20

-15

-10 1 0

0 0 2 0

0 0 3 0

0 0 4 0

0 0 5 0

0 0 6 0

0 0 7 0

0 0 8 0

0 0 9 0

0 0 10 0 0

0 11 0 0

0 12 0 0

0 13 0 0

0 14 0 0

0 15 0 0

0 16 0 0

0 17 0 0

0 18 0 0

0 19 0 0

0 20 0 0

0 21 0 0

0 22 0 0

0 23 0 0

0 24 0 0

0 TOTAL 0 0

0 5 0 5 10 15 -5 0 5 10 15 20 0 0

0 0

2 1 0 0

0 0

2 2 0 0

0 0

2 4 0 0

0 0

1 7 0 0

0 0 1 9 0 0 0 0 1

8 0 0

0 0

1 9 0 0

0 0

2 8 0 0

0 0

1 5 0 0

0 0

1 3 0 0

0 0

1 3 0 0

0 0

0 3 0 0

0 0

0 3 0 0

0 0

0 3 0 0

0 0

0 2 0 0

0 0

0 2 0 0

0 0

0 2 0 0

0 0

1 1 0 0

0 0

1 1 0 0

0 0

1 3 0 0

0 0

1 4 0 0

0 0

2 2 0 0

0 0

2 2 0 0

0 0

2 1 0 0

0 0 25 88 20 25 30 35 40 45 25 30 35 40 45 50 14 35 44 54 73 43 15 33 51 51 73 44 18 35 46 58 65 52 14 42 47 57 65 54 14 41 50 60 61 51 16 43 45 67 56 47 18 41 42 60 53 51 15 34 37 51 55 47 9 30 32 44 52 39 8 24 33 41 51 49 8 14 39 45 52 51 6 10 43 39 50 54 6 7 33 50 45 52 6 7 26 49 53 50 6 7 24 51 51 52 5 10 28 50 51 48 7 11 31 50 48 47 8 14 35 52 44 49 10 18 41 48 48 52 9 18 48 46 50 55 8 23 47 48 62 43 10 26 55 44 56 52 9 35 47 56 57 44 12 39 44 49 65 52 251 597 968 1220 1336 1178 50 55 60 65 70 75 55 60 65 70 75 80 48 39 8 0

0 0 50 33 6 1

0 0 49 27 5 0

0 0 48 23 3 0

0 0 51 18 4 0

0 0 51 23 3 0

0 0 38 39 8 0

0 0 39 41 18 0 0

0 43 36 23 3 0

0 42 38 33 5 0

0 39 49 39 6 0

0 44 50 43 8 0

0 49 50 46 12 0 0 49 52 47 15 0 0 52 47 45 21 0 0 51 46 44 24 0 0 51 44 45 23 0 0 43 51 44 17 0 0 42 46 41 14 0 0 49 39 35 9 0

0 55 45 22 4 0

0 48 46 19 1 0

0 49 48 11 1 0

0 44 44 9 0

0 0 1124 974 601 164 0 0 80 UNKNO TOTAL 0 4 365 0 4 365 0 4 365 0 4 365 0 5 365 0 5 365 0 5 365 0 18 365 0 48 365 0 37 365 0 19 365 0 15 365 0 12 365 0 8 365 0 7 365 0 6 365 0 6 365 0 6 365 0 3 365 0 3 365 0 3 365 0 4 365 0 4 365 0 4 365 0 234 8760 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000 LDC N-9 9-0 0 0 2.3-84 Table 2.3-19 Frequency of Occurrence, Dew Poi n t Temperature

( F) Versus Time of Day from 4

/74 through 3/75 for 33-ft Level T I ME O F DAY -40

-35 -40

-35

-30 1 0

0 0 2 0

0 0 3 0

0 0 4 0

0 0 5 0

0 0 6 0

0 0 7 0

0 0 8 0

0 0 9 0

0 0 10 0 0

0 11 0 0

0 12 0 0

0 13 0 0

0 14 0 0

0 15 0 0

0 16 0 0

0 17 0 0

0 18 0 0

0 19 0 0

0 20 0 0

0 21 0 0

0 22 0 0

0 23 0 0

0 24 0 0

0 T O T A L 0 0

0-30

-25

-20

-15 5 -25

-20

-15 5 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0 0 0 0 0 0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 5 10 15 20 25 5 10 15 20 25 30 0 1

4 12 23 45 0 2

2 14 22 47 0 1

5 12 24 44 0 2

4 13 22 46 0 2

4 13 28 43 0 3

2 16 23 44 0 2

3 16 23 46 1 1

2 16 22 48 1 1

3 8 17 48 1 1

4 5 18 48 1 2

3 7 21 44 1 3

3 10 20 44 1 1

4 11 20 48 1 3

2 11 23 47 1 3

3 11 25 48 1 2

4 10 26 49 1 2

3 11 28 49 0 2

4 13 24 45 0 1

5 11 22 47 0 1

4 10 25 47 0 1

2 12 26 45 0 1

2 13 30 43 0 1

2 16 24 49 0 1

3 13 28 48 10 40 77 284 564 11 1 2 30 35 40 45 50 55 35 40 45 50 55 60 74 86 67 38 7 3 70 77 77 38 8 3 74 74 73 42 9 2 67 77 73 44 10 2 65 81 65 48 6 4 68 77 66 46 11 2 60 78 64 42 21 4 48 81 59 44 20 4 47 70 56 47 16 4 45 74 65 48 15 5 53 87 56 54 15 4 56 89 66 47 9 3 59 90 67 42 8 3 69 80 73 40 6 2 66 78 78 38 6 1 66 90 73 30 7 1 66 92 69 29 6 3 74 86 71 24 9 3 76 86 74 27 10 3 73 88 70 30 10 3 76 91 68 28 6 4 76 91 62 33 8 1 70 88 65 35 7 3 73 78 67 40 6 4 15 7 1 19 9 1 16 2 6 934 238 71 60 65 70 75 80 65 70 75 80 UNKNO 1 0

0 0

0 4 1 0

0 0

0 4 1 0

0 0

0 4 1 0

0 0

0 4 1 0

0 0

0 5 2 0

0 0

0 5 1 0

0 0

0 5 1 0

0 0

0 18 0 0

0 0

0 47 0 0

0 0

0 36 0 0

0 0

0 18 0 0

0 0

0 14 0 0

0 0

0 11 0 0

0 0

0 8 0 0

0 0

0 7 0 0

0 0

0 6 0 0

0 0

0 6 0 0

0 0

0 6 0 0

0 0

0 3 1 0

0 0

0 3 1 0

0 0

0 3 1 0

0 0

0 4 1 0

0 0

0 4 0 0

0 0

0 4 13 0 0

0 0 229 TOTAL 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 87 6 0 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-85 Table 2.3-20 Monthly Averages of Psy c hrometric Data Based on P e riod of Record (1950-1970)

AVERAGES JAN FEB MAR APR MAY JUNE JULY AUG SEPT OCT NOV DEC YEAR DRY BULB WET BULB REL HUM.

DEWPOINT 30.3 27.9 76.0 23.2 37.5 33.6 69.7 27.4 44.0 37.3 55.0 27.3 52.5 42.8 46.4 30.4 61.8 49.1 41.8 36.0 69.9 54.5 39.4 41.2 77.5 57.9 31.5 42.3 75.3 57.3 34.9 42.8 67.0 52.6 39.9 39.5 53.2 45.4 57.7 36.9 40.1 36.4 72.6 31.1 33.4 31.2 80.8 27.5 53.5 43.8 53.8 33.8 MONTHLY AVERAGE EXTREMES

DRY BULB HIGHEST YEAR LOWEST YEAR 43.0 1953 12.9 1950 44.0 1958 25.8 1956 48.7 1968 39.6 1955 56.2 1956 48.3 1955 68.7 1958 57.2 1962 75.5 1969 64.2 1953 82.8 1960 73.2 1963 82.5 1967 70.6 1964 72.0 1967 61.6 1970 59.1 1952 50.3 1968 45.8 1954 32.3 1955 38.8 1953 26.5 1964 56.3 1953 51.0 1955+

WET BULB HIGHEST YEAR LOWEST YEAR 39.3 1953 12.4 1950 40.7 1958 23.4 1956 40.8 1963 32.9 1955 45.1 1962 39.3 1955 54.6 1958 45.4 1959 58.6 1958 51.4 1954 61.2 1958 55.6 1954 61.1 1961 54.9 1964 56.5 1963 48.3 1970 47.7 1962 42.4 1970 42.3 1954 29.6 1955 35.8 1966 25.0 1964 46.5 1958 41.8 1955 REL HUM. HIGHEST YEAR LOWEST YEAR 89 1960 60 1963 87 1963 54 1967 66 1950 44 1965 64 1963 37 1966 *52 1962+

31 1966 54 1950 34 1960 40 1955 22 1959 44 1968 24 1967 55 1969 34 1952 74 1962 42 1952 80 1956 64 1963+ 90 1950 69 1968 58 1950+

49 1967 DEWPOINT HIGHEST YEAR LOWEST YEAR 34.4 1953 6.5 1950 36.7 1958 17.3 1956 34.0 1961 20.8 1965+ 37.1 1963 26.2 1955 43.8 1957 30.4 1964 47.5 1958 37.5 1954 46.6 1958 35.4 1959 46.9 1961 38.4 1955 45.4 1963 33.8 1970 43.5 1962 32.1 1970 38.3 1954 24.0 1959 34.3 1950 21.0 1951 37.7 1958 31.5 1955 +Also in Earl ier Years *Although not included in these tables, an average of 63% was recorded in 1948 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-86 Table 2.3-21 Diurnal Variation of Preci p itation Intensity (Inches/Hour) at CGS and M onthly Total Precipi t a ti o n (Inches) at the Hanford Reservati o n HOUR APRIL 1974 MAY 1974 JUNE 1974 JULY 1974 AUGU ST 1974 SEPTE M BER 1974 OCTOBER 1974 NOVEMBER 1974 DECE M BER 1974 JANUARY 1975 FEB R UA R Y 1975 MA RCH 1975 AVERAGE F O R HOUR (A PRIL 1974 - M A RCH 1975) 01 02 03 04 05 06 07 08 09 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 0.000 0.000 0.016 0.000 0.016 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.024 0.032 0.016 0.016 0.024 0.016 0.020 0.000 0.000 0.032 0.080 0.036 0.136 0.016 0.000 0.000 0.032 0.048 0.072 0.000 0.088 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.048 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.024 0.032 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.168 0.040 0.032 0.016 0.000 0.016 0.032 0.000 0.000 0.000 0.016 0.000 0.000 0.000 0.000 0.016 0.056 0.000 0.024 0.000 0.000 0.000 0.032 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.064 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.016 0.000 0.064 0.000 0.000 0.000 0.016 0.000 0.016 0.000 0.000 0.000 0.000 0.016 0.024 0.056 0.024 0.000 0.000 0.000 0.000 0.032 0.036 0.040 0.028 0.052 0.028 0.000 0.000 0.000 0.000 0.016 0.040 0.032 0.020 0.020 0.016 0.016 0.000 0.000 0.000 0.000 0.024 0.056 0.020 0.036 0.024 0.032 0.000 0.040 0.000 0.032 0.000 0.024 0.048 0.000 0.000 0.020 0.016 0.064 0.072 0.020 0.024 0.024 0.000 0.056 0.024 0.036 0.024 0.032 0.048 0.040 0.032 0.032 0.020 0.024 0.016 0.048 0.024 0.032 0.032 0.000 0.000 0.000 0.000 0.016 0.024 0.016 0.072 0.040 0.016 0.016 0.000 0.000 0.024 0.064 0.036 0.000 0.024 0.016 0.000 0.020 0.032 0.040 0.000 0.000 0.000 0.000 0.000 0.016 0.048 0.032 0.024 0.028 0.024 0.024 0.000 0.000 0.000 0.000 0.000 0.000 0.048 0.088 0.000 0.016 0.080 0.071 0.029 0.026 0.018 0.030 0.045 0.026 0.029 0.039 0.036 0.034 0.022 0.032 0.024 0.033 0.029 0.038 0.026 0.034 0.032 0.040 0.021 0.040 0.038 MONTHLY TOTAL PR E C I P ITATION (INCHE S)

CGS H M S HMS (hist) 1946-19 7 0 Mean Total 0.55 0.46 0.44 0.44 0.28 0.50 0.06 0.12 0.66 0.45 0.71 0.16 0.00 trace 0.21 0.06 0.01 0.30 0.10 0.21 0.61 0.56 0.71 0.80 0.67 0.97 0.81 0.93 1.43 0.97 0.67 0.98 0.58 0.52 0.33 0.38 4.92 6.21 6.53 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-87 Table 2.3-22 Frequency of Occurrence, Precipitation (Inches/Hour) Versus Time of Day from 4/74 through 3/75 at CGS

Time of Day 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24TOTAL 016 6 5 7 4 5 5 8 5 6 4 5 5 6 6 8 7 9 8 7 8 4 3 8 10149 050 2 0 0 0 1 1 1 0 2 1 1 0 1 0 2 0 2 1 1 1 1 0 2 121 100 2 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 03 250 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 00 500 6 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 00

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-88 Table 2.3-22a Annual Frequency of Occu rrence of Wind Direction and Wind Speed Versus Precipitation Intensity FREQUENCY OF OCCURRENCE, WIND DIRECTION VS SPEED FROM 4/74 THROUGH 3/75 AT WPPSS2 FOR 33 FOOT LEVEL RAIN INTENSITY GREATER THAN OR EQUAL .016 INCHES PER HOUR SPEED CLASS (MPH)

NNE NE ENE CALM 0 0 0 1-3 2 2

2 4-7 2 3

1 8-12 0 0

0 13-18 0 0

0 19-24 0 0

0 25-UP 0 0

0 UNKNO 0 0

0 TOTAL 4 5

3 E ESE SE SSE S SSW 0 0

0 0

0 0 1 7

1 4

3 3 4 2

1 6

3 3 0 0

2 4

5 4 0 0

1 4

0 3 0 0

0 1

0 1 0 0

0 0

0 0 0 0

0 0

0 0 5 9

5 21 11 14 SW WSW W WNW NW NNW 0 0

0 0

0 0 1 1

0 0

5 2 3 2

3 5 10 10 4 1

2 7

5 2 1 0

0 0

1 0 0 1

0 0

0 0 0 0

1 0

0 0 0 0

0 0

0 0 9 5

6 12 21 14 N VAR CALM UNKNO TOTAL 0 0

0 0

0 0 2

0 0 36 1 1

0 0 60 0 0

0 0 36 0 0

0 0 10 0 0

0 0

3 0 0

0 0

1 0 0

0 1

1 1 3

0 1 149 FREQUENCY OF OCCURRENCE, WIND DIRECTION VS SPEED FROM 4/74 THROUGH 3/75 AT WPPSS2 FOR 33 FOOT LEVEL RAIN INTENSITY GREATER THAN OR EQUAL .050 INCHES PER HOUR SPEED CLASS (MPH)

NNE NE ENE CALM 0 0

0 1-3 1 1

0 4-7 1 0

0 8-12 0 0

0 13-18 0 0

0 19-24 0 0

0 25-UP 0 0

0 UNKNO 0 0

0 TOTAL 2 1

0 E ESE SE SSE S SSW 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

1 3

0 2 0 0

0 2

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

1 5

0 2 SW WSW W WNW NW NNW 0 0

0 0 0 0 0 0

0 0 0 0 0 0

0 2 1 1 0 1

1 0 3 1 0 0

0 0 0 0 0 0

0 0 0 0 0 0

0 0 0 0 0 0

0 0 0 0 0 1

1 2 4 2 N VAR CALM UNKNO TOTAL 0 0

0 0

0 0 0

0 0

2 0 0

0 0

5 0 0

0 0 12 0 0

0 0

2 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0 21 FREQUENCY OF OCCURRENCE, WIND DIRECTION VS SPEED FROM 4/74 THROUGH 3/75 AT WPPSS2 FOR 33 FOOT LEVEL RAIN INTENSITY GREATER THAN OR EQUAL .100 INCHES PER HOUR SPEED CLASS (MPH)

NNE NE ENE CALM 0 0

0 1-3 0 0

0 4-7 0 0

0 8-12 0 0

0 13-18 0 0

0 19-24 0 0

0 25-UP 0 0

0 UNKNO 0 0

0 TOTAL 0 0

0 E ESE SE SSE S SSW 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

1 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 1

0 0 SW WSW W WNW NW NNW 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 1

0 0 0 0

0 0

0 1 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 1

0 1 N VAR CALM UNKNO TOTAL 0 0

0 0

0 0 0

0 0

0 0 0

0 0

1 0 0

0 0

1 0 0

0 0

1 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

3 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-89 Table 2.3-22a Annual Frequency of Occurrence of Wind Direction and Wind Speed Versus Precipitation In tensity (Continued)

FREQUENCY OF OCCURRENCE, WIND DIRECTION VS SPEED FROM 4/74 THROUGH 3/75 AT WPPSS2 FOR 33 FOOT LEVEL RAIN INTENSITY GREATER THAN OR EQUAL .250 INCHES PER HOUR SPEED CLASS (MPH)

NNE NE ENE CALM 0 0

0 1-3 0 0

0 4-7 0 0

0 8-12 0 0

0 13-18 0 0

0 19-24 0 0

0 25-UP 0 0

0 UNKNO 0 0

0 TOTAL 0 0

0 E ESE SE SSE S SSW 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 SW WSW W WNW NW NNW 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 N VAR CALM UNKNO TOTAL 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 FREQUENCY OF OCCURRENCE, WIND DIRECTION VS SPEED FROM 4/74 THROUGH 3/75 AT WPPSS2 FOR 33 FOOT LEVEL RAIN INTENSITY GREATER THAN OR EQUAL .500 INCHES PER HOUR SPEED CLASS (MPH)

NNE NE ENE CALM 0 0 0 1-3 0 0

0 4-7 0 0

0 8-12 0 0

0 13-18 0 0

0 19-24 0 0

0 25-UP 0 0

0 UNKNO 0 0

0 TOTAL 0 0

0 E ESE SE SSE S SSW 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 SW WSW W WNW NW NNW 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 0 0

0 0

0 0 N VAR CALM UNKNO TOTAL 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 0 0

0 0

0 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-90 Table 2.3-23 Statistic on Fog at the Hanford M e teo r ology S t ation (Based on 1945-1970 Data)*

All Fog (viz 0-6 Miles)

Den s e Fog (viz 1/4 Mile or Less)

No. of Days No. of Hours No. of Days No. of Hours Great e s t No. of Hours of

Persis t ence J

F M

A M

J J

A S

O N

D Avg 9 6

1

2 8 12 Great e s t 19 20 6 1

3 1

1 1

1 9 14 20 Le ast 0 0

0 0

0 0

0 0

0 0

1 2 Avg 68.3 36.4 4.4 0.3 0.3

0.3 7.6 55.4 105.4 Great e s t 193.4 206.2 20.6 2.8 2.7 0.5 0.7 1.0 5.5 63.6 148.0 193.8 Le ast 0 0

0 0

0 0

0 0

0 0

1.0 6.5 Avg 6 3

1

0 0

1 5

8 Great e s t 14 11 5 1

1 0

0 1

1 6 13 17 Le ast 0 0

0 0

0 0

0 0

0 0

0 2 Avg 20.4 12.7 1.8 0.1 0.1 0

0

0.1 3.1 21.1 42.0 Great e s t 52.4 86.7 7.8 1.8 1.6 0

0 1.0 3.2 35.2 71.4 119.8 Le ast 0 0

0 0

0 0

0 0

0 0

0 1.3 (9) AF 58.1 58.0 12.2 2.8 2.7 0.5 0.7 0.7 2.6 39.0 65.4 72.3 (10) DF 15.0 16.7 5.0 0

1.6 0

0 0.7 1.4 15.8 20.6 47.0 (1) (2) (3) (4) (5) (6) (7) (8) Y 33 57 22 278.4 462.5 147.7 24 42 2 101.4 201.5 24.3 72.3 47.0 # Less than 1/2 1. Greatest number of days in a season -- occurred in 1969-70 2. Least number of days in a season -- occurred in 1948-49

3. Greatest number of hours in a season -- occurred in 1964-65

4. Least number of hours in a season -- occurred in 1948-49

5. Greatest number of days in a season -- occurred in 1950-51

6. Least number of days in a season -- occurred in 1948-49

7. Greatest number of hours in a season -- occurred in 1962-63

8. Least number of hours in a season -- occurred in 1948-49

9. AF denotes all fog (viz 0-6 miles)

10. DF denotes dense fog (viz 1/4 mile or less). Records for persistence of dense fog did not begin until 1953.

• Summation for the year does not necessarily reflect the summation of individual months.

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 2.3-91 Table 2.3-24 Percent Frequency Distri bution of Wind Speeds Duri ng Hourly Observations of Fog at Pasco (1966-19 7 0) and at HMS (1960-1970)

Speed Clas s* Station Calm 1-3 4-7 8-12 12 Total HM S (1) 29 44 25 2 0 100 Pasc o (2) 61 8 24 6 1 100

• Speed classes are in units of mph for HMS, and in units of knots for Pasco.

(1) Statistics for HMS are only for hour ly observations of fog restric ting visibility to 1/2 mile or less.

(2) Statistics for Pasco are for all hourly obs ervations of fog (visibility 0-6 miles).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-92 Table 2.3-25 Percent Frequency of Occurrence of Stability (T Distribution) at the Hanford Reservation (1)

Month/Year/Site Very Unstable Less Than -2.5 Unstable -2.5 to -1.5 t Range (°F/200 ft)

Neutral -1.5 to -0.5 Moderately Stable

-0.5 to +3.5 Very Stable Greater Than 3.5 4/74 CGS 4/74 HMS APRIL 1955-1970 HMS (hist) 0.14 3.61 5.10 15.69 25.28 20.88 27.78 26.94 30.22 43.33 40.83 39.19 12.36 3.19 4.61 5/74 CGS 5/74 HMS MAY 1955-1970 HMS (hist) 0.27 6.18 8.33 23.39 33.20 22.56 30.51 26.34 30.18 35.89 32.39 34.71 8.06 1.88 4.22 6/74 CGS 6/74 HMS JUNE 1955-1970 HMS (hist) 1.25 7.36 8.60 35.42 33.33 26.25 17.36 24.72 30.75 30.42 31.39 31.45 5.42 3.19 2.95 7/74 CGS 7/74 HMS JULY 1955-1970 HMS (hist) 0.00 6.72 8.74 28.23 30.78 26.31 25.81 28.90 27.69 27.02 30.51 33.42 14.11 3.09 3.84 8/74 CGS 8/74 HMS AUGUST 1955-1970 HMS (hist) 0.00 8.20 7.33 28.09 32.12 23.73 20.83 18.28 26.55 23.79 33.74 37.65 25.54 7.66 4.74 9/74 CGS 9/74 HMS SEPTEMBER 1955-1970 HMS (hist) 0.00 6.94 5.05 21.67 25.69 19.90 20.83 17.64 25.11 21.25 33.06 40.89 35.14 16.67 9.05 10/74 CGS 10/74 HMS OCTOBER 1955-1970 HMS (hist) 0.00 3.36 2.23 14.38 20.16 11.82 18.15 17.74 27.03 25.81 41.26 48.87 38.98 17.47 10.06 11/74 CGS 11/74 HMS NOVEMBER 1955-1970 HMS (hist) 0.00 0.00 0.76 4.58 5.00 6.82 28.06 30.42 31.74 52.22 57.22 53.37 14.44 5.83 7.30 12/74 CGS 12/74 HMS DECEMBER 1955-1970 HMS (hist) 0.00 0.13 0.40 1.75 5.24 4.35 22.72 22.72 36.53 56.18 60.89 50.98 18.15 11.02 7.74 (1)t at CGS is computed from 33 to 245 foot levels; at HMS, t is computed from 50 to 250 foot level.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-93 Table 2.3-25 Percent F r equency of Occurrence of Stability (T Distribution) at the Hanford Reservation (1) (Continued)

M onth/Year/Site Very Unstable Less Than -2.5 Uns t ab l e -2.5 to -1.5 t Range (° F/200 ft) Neu t ral -1.5 to -0.5 Modera tely S t a b le -0.5 to +3.5 Very Sta b le Grea t er Than 3.5 1/75 CGS 1/74 H M S (19 7 5 Not Available)

JANUARY 1 9 5 5-19 7 0 H M S (hist) 0.00 0.13 0.34 2.55 5.65 4.73 30.51 29.30 34.78 53.90 59.54 52.23 10.62 5.11 7.91 2/75 CGS 2/74 H M S (19 7 5 Not Available)

F EBRUARY 1 9 55-1 9 70 H M S (hist) 0.00 0.30 1.51 8.78 14.43 9.29 30.51 26.93 28.24 49.40 51.19 52.05 10.57 7.14 8.90 3/75 CGS 3/74 H M S (19 7 5 Not Available)

M ARCH 1955-1970 H M S (hist) 0.13 1.48 3.49 20.03 19.09 15.84 17.34 26.34 28.22 42.07 47.04 45.25 11.96 6.05 6.61 April 1974 -

M a rch 1975 CGS

1955-19 7 0 H M S (hist) 0.14 4.32 17.17 16.04 24.25 29.80 38.21 41.84 17.17 6.49 April 1975 -

M a rch 1976 C G S 0.59 21.31 21.85 37.20 17.51 (1)t at CGS is computed from 33 to 245 foot levels; at HMS, t is computed from 50 to 250 foot level.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT N o v e mb e r 1998 2.3-94 Table 2.3-26 Frequency of Occurrence T ( F/200 ft) Versus Time of D a y from 4/74 through 3/

7 5 at CGS between 245 and 33 ft Levels

Time of Day 1 2 3 4 5 678 9 1 0 1 1 1 2 1 3 1 415 1 6 1 7 1 8 1 9 2 0 2 1 2 2 2 324TOTALLT-2.5 0 0 0 0 0 000 0 032212 11000 0 000 12 GE-2.5 LT-1.5 0 0 0 0 0 0 3 43 98 137 156 175 213 217 196 148 90 2 710 0 00 0 1504 GE-1.5 LT-0.5 17 16 16 16 18 37 97 137 134 150 168 158 123 123 145 168 181 167 111 52 33 21 18 18 2124 GE-0.5 LT-3.5 216 202 193 197 187 188 177 131 76 40 20 16 14 14 14 42 85 156 203 236 235 242 238 225 3347 GE-3.5 124 141 148 143 150 133 84 36 9 511111 02 8 4 1 7 1 9 19510 211 6 1504UNKNO 8 6 8 9 10 7 4 1 8 4 8 3 3 1 7 1 31297 66796 6 776 269 TOTAL 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 365 8760 Average f o r Hour 3.008 3.261 2.714 .243 -1.104 -1.432 -1.502 -1.240 -.155 1.656 2.228 2.685 2.899 3.226 3.227 1.499

-.654 -1.303 -1.521 -1.441 -.835

.762 2.144 2.383 .883 C OLUMBIA G ENERATING S TATION Amendment53 F INAL S AFETY A NALYSIS R EPORTNovember 1998 2.3-95 Table 2.3-27 Frequency of Occurrence, Sigma (°) Versus Time of Day from 4/74 through 3/75 at CGS for 33 ft Level Time of Day GE22.5 1 111 2 110 3 105 4 104 5 104 6 112 7 96 8 117 9 124 10 139 11 144 12 152 13 152 14 153 15 137 16 109 17 62 18 44 19 52 20 51 21 79 22 88 23 107 24 98 TOTAL 2550 LT22.5 GE17.5 30 33 32 38 33 32342932353944333646 393220212836262741790 LT17.5 GE12.5 43 48 54 52 52 47544846587055686050 5950363439435147411205 LT 12.5 GE7.5 104 91 87 74 94 9711211481797179838094 1071441511131121059698882354 LT7.5 GE3.75 60 61 64 75 62 56583131181917132327 396592121112908668771365 LT3.75 GE2.1 0 5 7 5 4 6100010000 1111138585687LT2.1 8 8 8 7 5 5020000000 00037524569 UNKNO 9 9 8 10 11 10102451362118161311 111111888899340TOTAL 365 365 365 365 365 365365365365365365365365365365 3653653653653653653653653658760 Average for Hour 22.406 23.575 22.42123.03127.39027.76828.29821.89514.19814.65618.66720.99723.638 22.506 22.794 21.73027.37828.12428.87427.136 17.01115.28517.97820.98522.384 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 2.3-96 Table 2.3-28 Joint Frequency Distribution of Wind Speed and Direction

PLANT NAME: COLUMBIA GENERATING STATION METEOROLOGICAL INSTRUMENTATION DATA PERIOD: JFD 1996-1999 WIND SENSORS HEIGHT: 10.0 METER TYPE OF RELEASE: GROUND LEVEL RELEASE DELTA-T HEIGHTS: 10 - 75 METERS SOURCE OF DATA: CGS ONSITE MET DATA TAKEN FROM FRAMATOME JFD FILES FOR 96-99 PROGRAM: PAVAN, 10/76, 8/79 REVISION, IMPLEMENTATION OF REGULATORY GUIDE 1.145, REVISION 1

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS A WIND SPEED (M/S)

TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL

.42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 TOTAL .42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 .029 .101 .231 .190 .168 .114 .051 .079 .003 .000 .000 .000 .96 .035 .123 .190 .145 .117 .088 .028 .057 .003 .006 .000 .000 .79 .027 .095 .161 .130 .060 .035 .016 .032 .006 .000 .000 .000 .56 .021 .073 .114 .035 .022 .006 .003 .000 .000 .000 .000 .000 .27 .024 .085 .095 .032 .016 .013 .000 .000 .000 .000 .000 .000 .26 .022 .079 .123 .095 .028 .013 .003 .000 .000 .000 .000 .000 .36 .025 .088 .247 .158 .139 .070 .019 .003 .000 .000 .000 .000 .75 .014 .051 .180 .247 .262 .114 .047 .022 .006 .003 .000 .000 .95 .024 .085 .224 .209 .180 .196 .180 .196 .063 .022 .000 .000 1.38 .017 .060 .155 .107 .107 .117 .079 .092 .016 .006 .003 .000 .76 .013 .044 .098 .073 .076 .073 .035 .066 .035 .003 .000 .000 .52 .016 .057 .063 .054 .038 .025 .013 .041 .025 .009 .000 .000 .34 .013 .047 .095 .028 .035 .025 .016 .028 .022 .009 .000 .000 .32 .020 .070 .098 .054 .016 .019 .019 .035 .019 .022 .000 .000 .37 .022 .079 .155 .085 .060 .013 .016 .019 .013 .009 .000 .000 .47 .024 .085 .269 .171 .085 .098 .054 .035 .000 .003 .003 .000 .83 .348 1.223 2.497 1.811 1.410 1.018 .578 .705 .212 .095 .006 .000 9.90 JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS B WIND SPEED (M/S)

TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL

.42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 TOTAL .42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 .007 .022 .063 .085 .047 .035 .019 .013 .003 .000 .000 .000 .29 .008 .025 .041 .057 .057 .032 .006 .006 .000 .006 .000 .000 .24 .001 .003 .035 .047 .032 .025 .003 .006 .000 .000 .000 .000 .15 .005 .016 .013 .013 .006 .003 .003 .009 .000 .000 .000 .000 .07 .007 .022 .019 .009 .006 .000 .000 .000 .000 .000 .000 .000 .06 .005 .016 .035 .016 .022 .006 .000 .000 .000 .000 .000 .000 .10 .006 .019 .038 .063 .038 .009 .006 .009 .000 .000 .000 .000 .19 .006 .019 .025 .076 .085 .051 .019 .013 .003 .000 .000 .000 .30 .005 .016 .070 .095 .076 .092 .047 .066 .009 .000 .000 .000 .48 .005 .016 .057 .025 .047 .047 .041 .060 .025 .013 .000 .000 .34 .003 .009 .051 .032 .022 .032 .016 .051 .025 .025 .000 .000 .27 .003 .009 .028 .016 .019 .013 .000 .006 .013 .000 .000 .000 .11 .004 .013 .028 .016 .019 .019 .028 .032 .003 .013 .000 .000 .17 .003 .009 .025 .022 .022 .019 .022 .016 .006 .013 .000 .000 .16 .008 .025 .038 .028 .038 .016 .013 .006 .003 .000 .000 .000 .18 .012 .038 .063 .070 .032 .025 .009 .013 .000 .000 .000 .000 .26 .092 .278 .629 .670 .569 .424 .234 .307 .092 .070 .000 .000 3.36 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 2.3-97 Table 2.3-28 Joint Frequency Distribution of Wind Speed and Direction (Continued)

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS C TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL

.42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 TOTAL .42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 .007 .019 .076 .079 .054 .051 .022 .022 .000 .000 .000 .000 .33 .011 .032 .060 .057 .047 .035 .009 .006 .006 .000 .000 .000 .26 .009 .025 .035 .035 .009 .016 .009 .006 .000 .000 .000 .000 .14 .005 .016 .013 .028 .003 .000 .000 .000 .000 .000 .000 .000 .07 .003 .009 .019 .022 .006 .000 .000 .000 .000 .000 .000 .000 .06 .004 .013 .009 .003 .013 .006 .000 .000 .000 .000 .000 .000 .05 .005 .016 .025 .063 .035 .022 .003 .016 .000 .000 .000 .000 .19 .002 .006 .047 .098 .076 .057 .035 .016 .000 .000 .000 .000 .34 .007 .019 .044 .082 .092 .063 .063 .047 .019 .006 .000 .000 .44 .008 .022 .057 .044 .044 .047 .057 .063 .032 .019 .000 .000 .39 .008 .022 .028 .028 .044 .047 .022 .054 .041 .009 .000 .000 .30 .003 .009 .044 .019 .022 .035 .016 .028 .044 .003 .000 .000 .22 .003 .009 .022 .028 .025 .022 .022 .032 .013 .003 .000 .000 .18 .003 .009 .028 .038 .009 .022 .022 .016 .028 .025 .000 .000 .20 .003 .009 .051 .054 .022 .022 .013 .009 .022 .000 .000 .000 .21 .007 .019 .054 .079 .051 .035 .022 .028 .003 .003 .000 .000 .30 .088 .256 .613 .759 .553 .480 .316 .345 .209 .070 .000 .000 3.69 JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS D WIND SPEED (M/S) TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL

.42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 TOTAL .42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 .035 .142 .332 .420 .218 .193 .130 .032 .003 .003 .000 .000 1.51 .025 .101 .215 .231 .186 .101 .047 .038 .025 .003 .000 .000 .97 .020 .082 .168 .171 .126 .107 .051 .013 .019 .013 .006 .000 .78 .018 .073 .104 .079 .060 .019 .000 .006 .000 .000 .000 .000 .36 .012 .051 .117 .073 .057 .016 .000 .000 .000 .000 .000 .000 .33 .010 .041 .101 .092 .082 .022 .016 .003 .000 .000 .000 .000 .37 .021 .085 .205 .193 .218 .133 .051 .013 .003 .000 .000 .000 .92 .018 .073 .322 .401 .395 .303 .136 .120 .009 .000 .000 .000 1.78 .025 .101 .322 .338 .398 .326 .228 .335 .123 .041 .006 .000 2.24 .015 .060 .243 .212 .205 .281 .240 .370 .215 .120 .022 .000 1.98 .031 .126 .136 .168 .161 .142 .107 .202 .142 .101 .016 .000 1.33 .019 .079 .123 .149 .079 .054 .082 .155 .133 .035 .000 .000 .91 .027 .111 .158 .092 .101 .126 .092 .107 .057 .016 .006 .000 .89 .030 .123 .158 .202 .202 .149 .171 .326 .161 .057 .006 .000 1.59 .028 .114 .471 .468 .414 .310 .231 .161 .088 .051 .006 .000 2.34 .031 .126 .405 .455 .408 .265 .149 .155 .047 .009 .009 .000 2.06 .363 1.489 3.581 3.742 3.312 2.547 1.729 2.035 1.027 .449 .079 .000 20.35 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 2.3-98 Table 2.3-28 Joint Frequency Distribution of Wind Speed and Direction (Continued)

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS E WIND SPEED (M/S) TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL

.42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 TOTAL .42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 .037 .142 .370 .307 .161 .079 .035 .025 .009 .003 .000 .000 1.17 .029 .111 .291 .250 .104 .032 .013 .041 .041 .000 .000 .000 .91 .019 .073 .199 .171 .145 .047 .009 .019 .022 .003 .000 .000 .71 .021 .082 .095 .085 .025 .028 .016 .003 .000 .000 .000 .000 .36 .018 .070 .070 .060 .016 .000 .000 .003 .000 .000 .000 .000 .24 .013 .051 .098 .063 .079 .038 .006 .000 .000 .000 .000 .000 .35 .025 .095 .247 .288 .348 .247 .101 .070 .003 .000 .000 .000 1.42 .029 .111 .411 .544 .670 .525 .288 .224 .009 .000 .000 .000 2.81 .051 .196 .490 .484 .484 .553 .430 .424 .076 .025 .006 .000 3.22 .043 .168 .392 .288 .196 .250 .316 .518 .348 .133 .025 .000 2.68 .045 .174 .281 .228 .199 .164 .199 .224 .196 .082 .016 .000 1.81 .043 .164 .307 .218 .142 .107 .104 .120 .038 .019 .000 .000 1.26 .051 .196 .310 .256 .272 .136 .136 .130 .038 .019 .000 .000 1.54 .074 .284 .446 .480 .455 .509 .442 .477 .202 .063 .006 .000 3.44 .066 .253 .651 .762 .610 .487 .389 .354 .098 .032 .000 .000 3.70 .045 .174 .676 .540 .414 .234 .107 .180 .066 .013 .000 .000 2.45 .607 2.342 5.332 5.022 4.320 3.436 2.592 2.813 1.147 .392 .054 .000 28.06 JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS F WIND SPEED (M/S)

TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL

.42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 TOTAL .42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 .075 .272 .531 .307 .041 .006 .009 .000 .000 .000 .000 .000 1.24 .055 .199 .446 .205 .025 .003 .000 .000 .000 .000 .000 .000 .93 .037 .136 .221 .126 .070 .009 .000 .000 .000 .000 .000 .000 .60 .016 .057 .095 .051 .032 .009 .000 .000 .000 .000 .000 .000 .26 .016 .057 .044 .025 .006 .000 .000 .000 .000 .000 .000 .000 .15 .014 .051 .076 .032 .013 .006 .000 .000 .000 .000 .000 .000 .19 .020 .073 .209 .174 .269 .111 .028 .013 .000 .000 .000 .000 .90 .035 .126 .512 .604 .582 .414 .092 .066 .006 .000 .000 .000 2.44 .055 .202 .629 .733 .578 .297 .130 .104 .009 .006 .000 .000 2.75 .041 .149 .455 .389 .288 .120 .107 .085 .047 .009 .000 .000 1.69 .049 .177 .303 .205 .117 .051 .032 .032 .003 .006 .000 .000 .97 .057 .209 .265 .183 .088 .051 .016 .016 .000 .003 .000 .000 .89 .065 .237 .322 .164 .142 .101 .051 .006 .003 .000 .000 .000 1.09 .070 .256 .442 .319 .326 .171 .088 .044 .000 .000 .000 .000 1.72 .083 .303 .733 .512 .442 .161 .063 .006 .000 .000 .000 .000 2.31 .072 .262 .670 .528 .177 .073 .013 .013 .000 .000 .000 .000 1.81 .759 2.765 5.954 4.558 3.195 1.583 .629 .386 .070 .025 .000 .000 19.92 C OLUMBIA G ENERATING S TATION Amendment59 F INAL S AFETY A NALYSIS R EPORTDecember 2007LDCN-05-009 2.3-99 Table 2.3-28 Joint Frequency Distribution of Wind Speed and Direction (Continued)

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS G WIND SPEED (M/S) TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL

.42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 TOTAL .42 1.01 2.03 3.02 4.00 5.03 6.01 8.02 10.04 13.03 18.04 44.70 .113 .332 .860 .281 .025 .000 .000 .000 .000 .000 .000 .000 1.61 .111 .326 .787 .237 .025 .000 .000 .000 .000 .000 .000 .000 1.49 .068 .199 .335 .180 .060 .009 .000 .000 .000 .000 .000 .000 .85 .030 .088 .092 .044 .019 .006 .003 .000 .000 .000 .000 .000 .28 .020 .060 .019 .003 .003 .000 .000 .000 .000 .000 .000 .000 .11 .022 .063 .019 .000 .000 .000 .000 .000 .000 .000 .000 .000 .10 .027 .079 .139 .114 .028 .009 .000 .000 .000 .000 .000 .000 .40 .054 .158 .357 .493 .307 .136 .032 .006 .003 .000 .000 .000 1.55 .059 .174 .408 .496 .313 .085 .028 .009 .006 .000 .000 .000 1.58 .047 .139 .335 .155 .111 .025 .019 .003 .000 .003 .000 .000 .84 .067 .196 .237 .082 .047 .000 .000 .000 .000 .000 .003 .000 .63 .061 .180 .120 .051 .019 .013 .003 .000 .000 .000 .000 .000 .45 .067 .196 .171 .047 .041 .016 .000 .000 .000 .000 .000 .000 .54 .071 .209 .307 .139 .044 .016 .000 .000 .000 .000 .000 .000 .79 .094 .275 .667 .395 .221 .019 .006 .000 .000 .000 .000 .000 1.68 .134 .392 .796 .405 .101 .003 .000 .000 .000 .000 .000 .000 1.83 1.046 3.066 5.648 3.123 1.365 .338 .092 .019 .009 .003 .003 .000 14.71 WIND MEASURED AT 10.0 METERS WIND SPEED CORRECTED TO THE RELEASE HEIGHT OF 10.0 METERS.

OVERALL WIND DIRECTION FREQUENCY WIND DIRECTION: N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW FREQUENCY: 7.1 5.6 3.8 1.7 1.2 1.5 4.8 10.2 12.1 8.7 5.8 4.2 4.7 8.3 10.9 9.5 OVERALL WIND SPEED FREQUENCY AS MEASURED ON THE TOWER:

MAX. WIND SPEED (M/S): .425 1.006 2.034 3.018 4.001 5.029 6.013 8.024 10.036 13.031 18.038 44.704 WIND SPEED FREQUENCY: 3.30 11.42 24.25 19.68 14.73 9.83 6.17 6.61 2.77 1.10 .14 .00 BUILDING AND RELEASE CHARACTERISTICS:

RELEASE HEIGHT: 10.00 METERS MIXING VOLUME COEFFICIENT:

0.50 BUILDING CROSS-SECTIONAL AREA: 2861.00 SQUARE METERS THE FIVE-PERCENT-FOR-THE-ENTIRE-SITE /Q IS LIMITING.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-07-044 2.3-100Table 2.3-28A Joint Frequency Distribution of Wind Speed and Direction

PLANT NAME: COLUMBIA GENERATING STATION METEOROLOGICAL INSTRUMENTATION DATA PERIOD: 1984-1989, AVERAGE HOURLY DATA WIND SENSORS HEIGHT: 10.0 METER TYPE OF RELEASE: GROUND LVL RLS DELTA-T HEIGHTS: 245FT, 33FT SOURCE OF DATA: ANNUAL MET DATA COLLECTION 1984-1989 COMMENTS: DATA PROCESSED UNDER-THE SUPPLY SYSTEM QA PROGRAM: PAVAN, 10/76, 8/79 REVISION, IMPLEMENTATION OF REGULATORY GUIDE 1.145

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS A WIND SPEED (M/S)

TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 0.002 0.188 0.460 0.095 0.029 0.000 0.000 0.77 0.002 0.102 0.319 0.079 0.061 0.000 0.000 0.56 0.002 0.072 0.140 0.052 0.014 0.000 0.000 0.28 0.000 0.016 0.025 0.002 0.000 0.000 0.000 0.04 0.002 0.038 0.057 0.005 0.000 0.000 0.000 0.10 0.000 0.063 0.125 0.005 0.002 0.000 0.000 0.19 0.000 0.084 0.226 0.041 0.005 0.000 0.000 0.36 0.005 0.138 0.573 0.168 0.034 0.000 0.000 0.92 0.005 0.131 0.613 0.389 0.186 0.000 0.000 1.32 0.000 0.115 0.312 0.231 0.294 0.009 0.002 0.96 0.005 0.100 0.235 0.138 0.213 0.032 0.000 0.72 0.002 0.127 0.211 0.109 0.125 0.000 0.000 0.57 0.000 0.118 0.136 0.054 0.054 0.007 0.000 0.37 0.002 0.129 0.181 0.081 0.104 0.000 0.000 0.50 0.005 0.197 0.299 0.111 0.075 0.000 0.000 0.69 0.005 0.231 0.410 0.086 0.016 0.002 0.000 0.75 0.036 1.850 4.322 1.646 1.211 0.050 0.002 9.12 JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS B WIND SPEED (M/S)

TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 0.000 0.038 0.111 0.018 0.016 0.000 0.000 0.18 0.000 0.045 0.061 0.016 0.000 0.000 0.000 0.12 0.000 0.014 0.029 0.007 0.002 0.000 0.000 0.05 0.000 0.007 0.009 0.000 0.000 0.000 0.000 0.02 0.000 0.007 0.007 0.002 0.000 0.000 0.000 0.02 0.002 0.020 0.041 0.005 0.000 0.000 0.000 0.07 0.000 0.027 0.111 0.014 0.002 0.000 0.000 0.15 0.000 0.061 0.172 0.043 0.007 0.000 0.000 0.28 0.000 0.041 0.190 0.070 0.045 0.000 0.000 0.35 0.002 0.050 0.111 0.104 0.072 0.002 0.000 0.34 0.000 0.018 0.063 0.041 0.043 0.000 0.000 0.17 0.002 0.036 0.066 0.045 0.063 0.009 0.000 0.22 0.000 0.054 0.057 0.016 0.020 0.000 0.000 0.15 0.000 0.038 0.059 0.048 0.072 0.005 0.000 0.22 0.002 0.059 0.095 0.016 0.018 0.000 0.000 0.19 0.000 0.043 0.136 0.038 0.011 0.000 0.000 0.23 0.009 0.559 1.318 0.482 0.374 0.016 0.000 2.76 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-07-044 2.3-101Table 2.3-28A Joint Frequency Distribution of Wind Speed and Direction (Continued)

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS C TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 0.000 0.095 0.177 0.061 0.032 0.000 0.000 0.36 0.000 0.059 0.097 0.027 0.005 0.000 0.000 0.19 0.002 0.020 0.070 0.014 0.000 0.000 0.000 0.11 0.000 0.029 0.032 0.000 0.000 0.000 0.000 0.06 0.000 0.020 0.034 0.002 0.002 0.000 0.005 0.06 0.000 0.027 0.059 0.009 0.007 0.000 0.000 0.10 0.000 0.043 0.143 0.023 0.000 0.000 0.000 0.21 0.000 0.102 0.238 0.079 0.020 0.000 0.000 0.44 0.000 0.070 0.188 0.154 0.086 0.000 0.000 0.50 0.002 0.059 0.161 0.106 0.052 0.005 0.000 0.38 0.000 0.079 0.115 0.100 0.054 0.002 0.000 0.35 0.002 0.066 0.136 0.095 0.068 0.007 0.000 0.37 0.000 0.075 0.102 0.075 0.045 0.000 0.002 0.30 0.005 0.100 0.120 0.041 0.091 0.005 0.000 0.36 0.000 0.088 0.170 0.045 0.063 0.000 0.000 0.37 0.000 0.109 0.179 0.029 0.023 0.000 0.000 0.34 0.011 1.041 2.019 0.860 0.548 0.018 0.007 4.50 JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS D WIND SPEED (M/S) TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 0.014 0.469 0.623 0.211 0.109 0.000 0.000 1.42 0.011 0.267 0.387 0.081 0.052 0.000 0.000 0.80 0.005 0.147 0.238 0.038 0.029 0.000 0.000 0.46 0.002 0.081 0.095 0.002 0.011 0.000 0.000 0.19 0.000 0.093 0.140 0.016 0.011 0.000 0.000 0.26 0.002 0.204 0.231 0.048 0.007 0.000 0.000 0.49 0.005 0.213 0.441 0.072 0.032 0.000 0.000 0.76 0.005 0.287 0.847 0.244 0.068 0.000 0.000 1.45 0.002 0.337 0.885 0.414 0.220 0.002 0.000 1.86 0.011 0.283 0.627 0.475 0.387 0.032 0.000 1.82 0.002 0.263 0.414 0.290 0.283 0.011 0.005 1.27 0.016 0.333 0.340 0.201 0.226 0.005 0.002 1.12 0.005 0.355 0.407 0.231 0.177 0.005 0.002 1.18 0.016 0.448 0.586 0.349 0.387 0.032 0.007 1.82 0.011 0.559 0.894 0.240 0.213 0.020 0.005 1.94 0.014 0.534 0.996 0.263 0.170 0.009 0.000 1.99 0.120 4.874 8.152 3.176 2.381 0.115 0.020 18.84 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-07-044 2.3-102Table 2.3-28A Joint Frequency Distribution of Wind Speed and Direction (Continued)

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS E WIND SPEED (M/S) TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 0.016 0.808 0.754 0.125 0.043 0.000 0.000 1.75 0.009 0.521 0.552 0.070 0.050 0.000 0.000 1.20 0.014 0.328 0.494 0.027 0.005 0.000 0.000 0.87 0.014 0.134 0.158 0.034 0.000 0.000 0.000 0.34 0.000 0.154 0.168 0.011 0.002 0.000 0.000 0.34 0.020 0.285 0.412 0.018 0.000 0.000 0.000 0.74 0.011 0.421 0.840 0.127 0.054 0.000 0.000 1.45 0.020 0.521 1.720 0.392 0.170 0.000 0.000 2.82 0.011 0.738 1.684 0.632 0.396 0.009 0.000 3.47 0.020 0.700 1.252 0.580 0.552 0.059 0.002 3.16 0.036 0.616 0.912 0.369 0.292 0.016 0.000 2.24 0.009 0.718 0.810 0.274 0.177 0.002 0.011 2.00 0.014 0.786 0.998 0.441 0.235 0.011 0.000 2.49 0.054 1.035 1.648 0.706 0.480 0.016 0.000 3.94 0.034 1.184 1.813 0.496 0.303 0.007 0.000 3.84 0.043 1.050 1.410 0.310 0.152 0.000 0.000 2.97 0.326 9.997 15.627 4.611 2.911 0.120 0.014 33.61 JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS F WIND SPEED (M/S)

TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 0.016 0.752 0.441 0.000 0.000 0.000 0.000 1.21 0.011 0.552 0.274 0.005 0.000 0.000 0.000 0.84 0.005 0.278 0.285 0.009 0.002 0.000 0.000 0.58 0.014 0.131 0.104 0.009 0.000 0.000 0.000 0.26 0.005 0.100 0.070 0.005 0.000 0.000 0.000 0.18 0.009 0.190 0.152 0.005 0.000 0.000 0.000 0.36 0.007 0.333 0.573 0.066 0.036 0.000 0.000 1.01 0.009 0.550 1.288 0.240 0.079 0.005 0.000 2.17 0.018 0.629 1.320 0.242 0.079 0.000 0.000 2.29 0.011 0.557 0.643 0.179 0.052 0.005 0.005 1.45 0.027 0.537 0.430 0.070 0.020 0.002 0.000 1.09 0.020 0.494 0.432 0.034 0.023 0.000 0.000 1.00 0.029 0.552 0.405 0.088 0.018 0.000 0.000 1.09 0.034 0.772 0.815 0.195 0.050 0.000 0.000 1.87 0.014 1.050 1.148 0.113 0.045 0.000 0.018 2.39 0.023 1.021 0.747 0.025 0.000 0.000 0.014 1.83 0.251 8.498 9.128 1.284 0.405 0.011 0.036 19.61 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-07-044 2.3-103Table 2.3-28A Joint Frequency Distribution of Wind Speed and Direction (Continued)

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION ATMOSPHERIC STABILITY CLASS G WIND SPEED (M/S) TOWER RELEASE N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 TOTAL 0.27 1.34 3.13 5.36 8.05 10.73 11.18 0.016 0.790 0.312 0.007 0.002 0.000 0.032 1.16 0.005 0.686 0.292 0.002 0.009 0.000 0.032 1.03 0.007 0.303 0.240 0.029 0.005 0.000 0.032 0.62 0.005 0.174 0.084 0.005 0.000 0.000 0.023 0.29 0.005 0.091 0.023 0.000 0.000 0.000 0.000 0.12 0.011 0.111 0.059 0.007 0.000 0.000 0.000 0.19 0.014 0.213 0.272 0.011 0.000 0.000 0.000 0.51 0.007 0.385 0.869 0.075 0.014 0.000 0.000 1.35 0.009 0.392 0.645 0.050 0.011 0.011 0.000 1.12 0.014 0.310 0.310 0.014 0.007 0.005 0.000 0.66 0.016 0.267 0.174 0.011 0.000 0.000 0.000 0.47 0.014 0.287 0.104 0.005 0.002 0.000 0.000 0.41 0.007 0.299 0.115 0.005 0.000 0.000 0.000 0.43 0.009 0.473 0.267 0.020 0.005 0.000 0.000 0.77 0.020 0.654 0.453 0.018 0.000 0.000 0.000 1.15 0.027 0.822 0.450 0.000 0.000 0.000 0.007 1.31 0.183 6.257 4.670 0.258 0.054 0.016 0.125 11.56 WIND MEASURED AT 10.0 METERS WIND SPEED CORRECTED TO THE RELEASE HEIGHT OF 10.0 METERS. OVERALL WIND DIRECTION FREQUENCY WIND DIRECTION: N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW FREQUENCY: 6.9 4.7 3.0 1.2 1.1 2.1 4.5 9.4 10.9 8.8 6.3 5.7 6.0 9.5 10.6 9.4 OVERALL WIND SPEED FREQUENCY AS MEASURED ON THE TOWER: MAX. WIND SPEED (M/S): 0.268 1.341 3.129 5.364 8.047 10.729 11.176 WIND SPEED FREQUENCY: 0.94 33.08 45.23 12.32 7.88 0.35 0.20 BUILDING AND RELEASE CHARACTERISTICS:

RELEASE HEIGHT: 10.00 METERS MIXING VOLUME COEFFICIENT: 0.50 BUILDING CROSS-SECTIONAL AREA: 2766.00 SQUARE METERS THE FIVE-PERCENT-FOR-THE-ENTIRE-SITE /Q IS LIMITING.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-104

TABLE 2.3-29 THROUGH TABLE 2.3-32 DELETED C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-105 Table 2.3-33 Exclusion Area Boundary Accident /Q Desert Sigmas PLANT NAME: CGS METEOROLOGICAL INSTRUMENTATION DATA PERIOD: JFD 1996-1999 WIND SENSORS HEIGHT: 10.0 METERS TYPE OF RELEASE: GROUND LEVEL RELEASE DELTA-T HEIGHTS: 10 - 75 METERS SOURCE OF DATA: CGS ONSITE MET DATA TAKEN FROM FRAMATOME JFD FILES FOR 96-99 COMMENTS: input file: P96-99-F.inp output file: P96-99-F.out PROGRAM: PAVAN, 10/76, 8/79 REVISION, IMPLEMENTATION OF REGULATORY GUIDE 1.145, REVISION 1 RELATIVE CONCENTRATION (/Q) VALUES (SEC/CUBIC METER)

VERSUS HOURS PER YEAR MAX AVERAGING TIME 0-2 HR /Q IS DOWNWIND SECTOR DISTANCE (METERS) 0-2 HOURS

0-8 HOURS 8-24 HOURS 1-4 DAYS 4-30 DAYS ANNUAL AVERAGE EXCEEDED IN SECTOR DOWWIND SECTOR S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1950. 1.68E-04 1.61E-04 1.13E-04 5.62E-05 2.88E-05 3.14E-05 7.22E-05 1.19E-04 1.30E-04 1.09E-04 1.20E-04 1.14E-04 1.25E-04 1.39E-04 1.59E-04 1.81E-04 1.08E-04 1.02E-04 7.01E-05 3.38E-05 1.80E-05 1.94E-05 4.41E-05 7.70E-05 8.49E-05 6.85E-05 7.40E-05 6.94E-05 7.65E-05 8.76E-05 1.04E-04 1.18E-04 8.69E-05 8.16E-05 5.51E-05 2.63E-05 1.42E-05 1.53E-05 3.45E-05 6.20E-05 6.86E-05 5.45E-05 5.81E-05 5.41E-05 6.00E-05 6.94E-05 8.44E-05 9.52E-05 5.39E-05 5.00E-05 3.27E-05 1.51E-05 8.52E-06 9.06E-06 2.02E-05 3.87E-05 4.32E-05 3.31E-05 3.44E-05 3.16E-05 3.53E-05 4.20E-05 5.34E-05 5.98E-05 2.71E-05 2.48E-05 1.55E-05 6.88E-06 4.09E-06 4.29E-06 9.39E-06 1.97E-05 2.23E-05 1.62E-05 1.62E-05 1.46E-05 1.65E-05 2.04E-05 2.77E-05 3.07E-05 1.17E-05 1.05E-05 6.20E-06 2.62E-06 1.67E-06 1.72E-06 3.67E-06 8.64E-06 9.87E-06 6.73E-06 6.46E-06 5.66E-06 6.50E-06 8.41E-06 1.24E-05 1.36E-05 38.6 389.5 22.0 10.4 7.6 7.5 10.0 20.3 23.5 17.9 23.8 22.7 25.1 27.9 34.5 43.7 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE MAX /Q 1.81E-04 TOTAL HOURS AROUND SITE: 724.9 SRP 2.3.4 1950. 1.69E-04 1.12E-04 9.06E-05 5.76E-05 3.01E-05 1.36E-05 SITE LIMIT 1.69E-04 1.12E-04 9.06E-055.76E-05 3.01E-05 1.36E-05 THE FIVE-PERCENT-FOR-THE-ENTIRE-SITE /Q IS LIMITING.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-106 Table 2.3-33a Exclusion Area Boundary /Q Values Desert Sigmas w/ Meander

Direction From Site 0.5% Level(a) (10-4 sec/m 3) 5% Level (b) (10-4 sec/m 3) S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 1.68 1.61 1.13 0.562 0.288 0.314 0.722 1.19 1.30 1.09 1.20 1.14 1.25 1.39 1.59 1.81 2.04 2.26 2.13 2.20 2.19 1.93 1.17 1.18 1.17 1.18 1.74 2.03 2.00 1.58 1.52 1.86 (a) Exceeded 0.5% of the total time.

(b) Exceeded 5% of the time that wind blows into the individual sector.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-107 Table 2.3-34 Exclusion Area Boundary Accident /Q P-G Sigmas PLANT NAME: CGS METEOROLOGICAL INSTRUMENTATION DATA PERIOD: JFD 1996-1999 WIND SENSORS HEIGHT: 10.0 METERS TYPE OF RELEASE: GROUND LEVEL RELEASE DELTA-T HEIGHTS: 10 - 75 METERS SOURCE OF DATA: CGS ONSITE MET DATA TAKEN FROM FRAMATOME JFD FILES FOR 96-99 COMMENTS: input file: P96-99-F.inp output file: P96-99-F.out PROGRAM: PAVAN, 10/76, 8/79 REVISION, IMPLEMENTATION OF REGULATORY GUIDE 1.145, REVISION 1 RELATIVE CONCENTRATION (/Q) VALUES (SEC/CUBIC METER)

VERSUS HOURS PER YEAR MAX AVERAGING TIME 0-2 HR /Q IS DOWNWIND SECTOR DISTANCE (METERS) 0-2 HOURS 0-8 HOURS 8-24 HOURS

1-4 DAYS 4-30 DAYS ANNUAL AVERAGE EXCEEDED IN SECTOR DOWWIND SECTOR S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 1950.1950.1950.1950.1950.1950.1950.1950.1950.1950.1950.1950.1950.1950.1950.1950. 1.67E-041.59E-041.15E-045.48E-052.52E-052.85E-057.40E-051.30E-041.38E-041.08E-041.09E-041.01E-041.10E-041.24E-041.59E-041.77E-04 8.92E-058.36E-055.89E-052.81E-051.39E-051.54E-053.92E-057.05E-057.60E-055.86E-055.81E-055.32E-055.87E-056.83E-058.86E-059.68E-05 6.53E-056.06E-054.21E-052.01E-051.03E-051.13E-052.85E-055.20E-055.65E-054.31E-054.24E-053.87E-054.29E-055.07E-056.61E-057.16E-05 3.32E-053.01E-052.04E-059.75E-065.39E-065.80E-061.43E-052.68E-052.97E-052.22E-052.15E-051.94E-052.17E-052.65E-053.51E-053.72E-05 1.25E-051.10E-057.20E-063.45E-062.12E-062.22E-065.29E-061.04E-051.18E-058.56E-068.07E-067.19E-068.18E-061.05E-051.41E-051.45E-05 3.82E-063.23E-062.01E-069.65E-076.80E-076.87E-071.57E-063.24E-063.81E-062.67E-062.44E-062.14E-062.47E-063.37E-064.63E-064.59E-06 38.9 385.6 21.0 8.6 6.1 6.3 9.2 22.8 25.5 17.2 19.9 18.9 20.7 23.2 35.243.7 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE MAX /Q 1.77E-04 TOTAL HOURS AROUND SITE: 702.8 SRP 2.3.4 1950. 2.86E-04 1.45E-04 1.03E-04 4.91E-05 1.70E-05 4.63E-06 SITE LIMIT 1.65E-04 9.16E-05 6.82E-05 3.59E-05 1.43E-05 4.63E-06 THE FIVE-PERCENT-FOR-THE-ENTIRE-SITE /Q IS LIMITING.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-108 Table 2.3-34a Exclusion Area Boundary /Q Values Pasquill-Gifford Sigmar w/ M eander and Building Wake Credit Direction From Site 0.5% Level(a) (10-4 sec/m 3) 5% Level (b) (10-4 sec/m 3) S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 1.67 1.59 1.15 0.548 0.252 0.285 0.740 1.30 1.38 1.08 1.09 1.01 1.10 1.24 1.59 1.77 1.99 2.18 2.02 1.96 1.93 1.71 1.14 1.29 1.25 1.17 1.53 1.80 1.77 1.38 1.52 1.82 (a) Exceeded 0.5% of the total time.

(b) Exceeded 5% of the time that wind blows into the individual sector.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-109 Table 2.3-35 Low Population Zone Accident /Q Desert Sigmas PLANT NAME: CGS METEOROLOGICAL INSTRUMENTATION DATA PERIOD: JFD 1996-1999 WIND SENSORS HEIGHT: 10.0 METERS TYPE OF RELEASE: GROUND LEVEL RELEASE DELTA-T HEIGHTS: 10 - 75 METERS SOURCE OF DATA: CGS ONSITE MET DATA TAKEN FROM FRAMATOME JFD FILES FOR 96-99 COMMENTS: input file: P96-99-F.inp output file: P96-99-F.out PROGRAM: PAVAN, 10/76, 8/79 REVISION, IMPLEMENTATION OF REGULATORY GUIDE 1.145, REVISION 1 RELATIVE CONCENTRATION (/Q) VALUES (SEC/CUBIC METER)

VERSUS HOURS PER YEAR MAX AVERAGING TIME 0-2 HR /Q IS DOWNWIND SECTOR DISTANCE (METERS) 0-2 HOURS 0-8 HOURS 8-24 HOURS 1-4 DAYS 4-30 DAYS ANNUAL AVERAGE EXCEEDED IN SECTOR DOWNWIND SECTOR S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 8.35E-05 7.96E-05 5.52E-05 2.18E-05 7.98E-06 8.94E-06 2.94E-05 5.22E-05 5.78E-05 5.01E-05 5.49E-05 5.04E-05 5.66E-05 6.26E-05 7.62E-05 8.91E-05 4.58E-05 4.33E-05 2.92E-05 1.16E-05 4.64E-06 5.13E-06 1.57E-05 2.93E-05 3.26E-05 2.71E-05 2.92E-05 2.66E-05 2.99E-05 3.39E-05 4.27E-05 4.95E-05 3.39E-05 3.19E-05 2.12E-05 8.48E-06 3.54E-06 3.89E-06 1.15E-05 2.20E-05 2.45E-05 1.99E-05 2.12E-05 1.93E-05 2.17E-05 2.50E-05 3.20E-05 3.69E-05 1.76E-05 1.64E-05 1.06E-05 4.28E-06 1.97E-06 2.13E-06 5.81E-06 1.17E-05 1.31E-05 1.03E-05 1.07E-05 9.60E-06 1.09E-05 1.28E-05 1.70E-05 1.95E-05 6.92E-06 6.36E-06 3.94E-06 1.61E-06 8.46E-07 8.96E-07 2.19E-06 4.78E-06 5.38E-06 3.95E-06 3.98E-06 3.53E-06 4.02E-06 4.94E-06 6.91E-06 7.81E-06 2.20E-06 1.99E-06 1.17E-06 4.85E-07 3.02E-07 3.11E-07 6.61E-07 1.59E-06 1.80E-06 1.23E-06 1.19E-06 1.04E-06 1.19E-06 1.54E-06 2.29E-06 2.55E-06 39.1 364.8 22.2 9.8 7.0 7.2 9.1 18.0 21.1 16.7 22.7 21.6 23.7 25.3 33.4 43.7 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE MAX /Q 8.91E-05 TOTAL HOURS AROUND SITE: 685.7 SRP 2.3.4 4827. 7.96E-05 4.51E-05 3.39E-05 1.83E-05 7.54E-06 2.55E-06 SITE LIMIT 7.96E-05 4.51E-05 3.39E-05 1.83E-05 7.54E-06 2.55E-06 THE FIVE-PERCENT-FOR-THE-ENTIRE-SITE /Q IS LIMITING.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-110 Table 2.3-36 Low Population Zone Accident /Q P-G Sigmas PLANT NAME: CGS METEOROLOGICAL INSTRUMENTATION DATA PERIOD: JFD 1996-1999 WIND SENSORS HEIGHT: 10.0 METERS TYPE OF RELEASE: GROUND LEVEL RELEASE DELTA-T HEIGHTS: 10 - 75 METERS SOURCE OF DATA: CGS ONSITE MET DATA TAKEN FROM FRAMATOME JFD FILES FOR 96-99 COMMENTS: input file: P96-99-F.inp output file: P96-99-F.out PROGRAM: PAVAN, 10/76, 8/79 REVISION, IMPLEMENTATION OF REGULATORY GUIDE 1.145, REVISION 1 RELATIVE CONCENTRATION (/Q) VALUES (SEC/CUBIC METER)

VERSUS HOURS PER YEAR MAX AVERAGING TIME 0-2 HR /Q IS DOWNWIND SECTOR DISTANCE (METERS) 0-2 HOURS 0-8 HOURS 8-24 HOURS 1-4 DAYS 4-30 DAYS ANNUAL AVERAGE EXCEEDED IN SECTOR DOWNWIND SECTOR S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 4827. 7.65E-05 7.30E-05 5.16E-05 2.17E-05 8.35E-06 9.39E-06 3.16E-05 5.63E-05 6.01E-05 4.77E-05 4.98E-05 4.56E-05 5.08E-05 5.64E-05 7.21E-05 8.17E-05 3.43E-05 3.21E-05 2.22E-05 9.52E-06 4.05E-06 4.48E-06 1.42E-05 2.59E-05 2.81E-05 2.18E-05 2.23E-05 2.03E-05 2.28E-05 2.61E-05 3.38E-05 3.74E-05 2.30E-05 2.13E-05 1.46E-05 6.31E-06 2.83E-06 3.10E-06 9.49E-06 1.76E-05 1.92E-05 1.48E-05 1.49E-05 1.35E-05 1.52E-05 1.78E-05 2.31E-05 2.53E-05 9.64E-06 8.74E-06 5.85E-06 2.59E-06 1.29E-06 1.39E-06 3.97E-06 7.56E-06 8.42E-06 6.32E-06 6.22E-06 5.62E-06 6.37E-06 7.71E-06 1.02E-05 1.08E-05 2.77E-06 2.43E-06 1.57E-06 7.19E-07 4.19E-07 4.38E-07 1.14E-06 2.25E-06 2.58E-06 1.87E-06 1.78E-06 1.59E-06 1.82E-06 2.32E-06 3.12E-06 3.21E-06 6.01E-07 5.08E-07 3.16E-07 1.50E-07 1.06E-07 1.07E-07 2.46E-07 5.13E-07 6.05E-07 4.21E-07 3.84E-07 3.40E-07 3.94E-07 5.36E-07 7.35E-07 7.24E-07 38.9 379.7 20.9 9.2 6.5 6.7 9.1 21.4 24.0 16.8 21.1 20.1 22.1 23.6 34.6 43.7 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE MAX /Q 8.17E-05 TOTAL HOURS AROUND SITE: 698.5 SRP 2.3.4 4827. 1.15E-04 4.99E-05 3.28E-05 1.33E-05 3.61E-06 7.35E-07 SITE LIMIT 7.53E-05 3.50E-05 2.39E-05 1.04E-05 3.16E-06 7.35E-07 THE FIVE-PERCENT-FOR-THE-ENTIRE-SITE /Q IS LIMITING.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-009 2.3-111 Table 2.3-37 Control Room, Exclusion Area Boundary and Low Population Zone /Qs (S/m 3) Control Room (1) LPZ (2) EAB (2) Filtered Unfiltered SGT Roofline Railway Bay doors SC Leakage RBW SC Leakage Turbine Building SGT Roofline Railway Bay doors SC Leakage RBW SC Leakage Turbine Building 0 - 2 hrs 1.43E-04 3.65E-04 1.99E-04 8.81E-04 6.95E-04 5.34E-04 8.69E-04 4.70E-03 4.95E-05 1.81E-04 2 - 8 hrs 1.05E-04 2.89E-04 1.44E-04 3.75E-04 3.36E-04 1.97E-04 4.40E-04 2.00E-03 4.95E-05 8 - 24 hrs 4.14E-05 1.18E-04 5.73E-05 1.93E-04 1.28E-04 8.41E-05 1.75E-04 1.03E-03 3.69E-05 1 - 4 days 3.52E-05 9.83 E-05 5.00E-05 1.50E-04 9.72E-05 7.26E-05 1.38E-04 8.01E-04 1.95E-05 4 - 30 days 3.03E-05 8.61E-05 4.18E-05 1.44E-04 7.69E-05 7.00E-05 1.10E-04 7.69E-04 7.81E-06 (1) Reference 2.3-37 (2) Reference 2.3-38 NOTE: EAB = Exclusion Area Boundary LPZ = Low Population Zone SGT = Standby Gas Treatment SC = Secondary Containment RBW = Reactor Building Wall

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-112 Table 2.3-38a CGS Calculation, Terrain Features, Desert Sigmas CGS TURBINE AND RADWASTE BLDGS NO DECAY, UNDEPLETED CORRECTED USING STANDARD OPEN TERRAIN FACTORS ANNUAL AVERAGE CHI/Q (SEC/METER CUBED)

DISTANCE IN MILES FROM THE SITE SECTOR 0.250 0.500 0.750 1.000 1.500 2.000 2.500 3.000 3.500 4.000 4.500 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 1.170E-04 8.738E-05 5.041E-05 2.672E-05 1.743E-05 3.424E-05 5.972E-05 1.094E-04 1.148E-04 9.052E-05 7.851E-05 7.545E-05 8.001E-05 1.238E-04 1.508E-04 1.535E-04 4.880E-05 3.691E-05 2.119E-05 1.129E-05 7.139E-06 1.390E-05 2.452E-05 4.499E-05 4.680E-05 3.669E-05 3.196E-05 3.053E-05 3.243E-05 5.040E-05 6.203E-05 6.378E-05 2.943E-05 2.243E-05 1.282E-05 6.851E-06 4.257E-06 8.233E-06 1.463E-05 2.691E-05 2.783E-05 2.171E-05 1.896E-05 1.805E-05 1.917E-05 2.986E-05 3.703E-05 3.835E-05 1.612E-05 1.234E-05 7.028E-06 3.769E-06 2.314E-06 4.450E-06 7.962E-06 1.468E-05 1.514E-05 1.174E-05 1.029E-05 9.773E-06 1.038E-05 1.617E-05 2.016E-05 2.097E-05 7.145E-06 5.502E-06 3.117E-06 1.679E-06 1.015E-06 1.936E-06 3.496E-06 6.471E-06 6.639E-06 5.115E-06 4.493E-06 4.260E-06 4.517E-06 7.040E-06 8.853E-06 9.262E-06 4.144E-06 3.204E-06 1.808E-06 9.761E-07 5.849E-07 1.109E-06 2.014E-06 3.738E-06 3.821E-06 2.930E-06 2.578E-06 2.442E-06 2.585E-06 4.029E-06 5.097E-06 5.358E-06 2.767E-06 2.146E-06 1.207E-06 6.527E-07 3.886E-07 7.331E-07 1.337E-06 2.487E-06 2.536E-06 1.937E-06 1.706E-06 1.615E-06 1.708E-06 2.662E-06 3.383E-06 3.569E-06 2.013E-06 1.565E-06 8.781E-07 4.753E-07 2.817E-07 5.293E-07 9.689E-07 1.805E-06 1.836E-06 1.399E-06 1.233E-06 1.167E-06 1.232E-06 1.920E-06 2.449E-06 2.593E-06 1.551E-06 1.208E-06 6.766E-07 3.666E-07 2.164E-07 4.053E-07 7.440E-07 1.388E-06 1.409E-06 1.071E-06 9.443E-07 8.935E-07 9.420E-07 1.469E-06 1.879E-06 1.995E-06 1.246E-06 9.718E-07 5.432E-07 2.945E-07 1.733E-07 3.237E-07 5.955E-07 1.112E-06 1.128E-06 8.551E-07 7.543E-07 7.136E-07 7.514E-07 1.172E-06 1.503E-06 1.600E-06 1.031E-06 8.058E-07 4.497E-07 2.439E-07 1.431E-07 2.667E-07 4.917E-07 9.193E-07 9.305E-07 7.045E-07 6.216E-07 5.881E-07 6.185E-07 9.644E-07 1.241E-06 1.323E-06 ANNUAL AVERAGE CHI/Q (SEC/METER CUBED) DISTANCE IN MILES FROM THE SITE SECTOR 5.000 7.500 10.000 15.000 20.000 25.000 30.000 35.000 40.000 45.000 50.000

S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 8.745E-07 6.840E-07 3.812E-07 2.068E-0 7 1.211E-07 2.252E-07 4.159E-07 7.782E-07 7.867E-07 5.948E-07 5.248E-07 4.965E-07 5.217E-07 8.135E-07 1.049E-06 1.120E-06 4.911E-07 3.859E-07 2.140E-07 1.162E-07 6.750E-08 1.245E-07 2.313E-07 4.434E-07 4.368E-07 3.285E-07 2.900E-07 2.744E-07 2.872E-07 4.479E-07 5.821E-07 6.263E-07 3.377E-07 2.662E-07 1.471E-07 7.994E-08 4.618E-08 8.474E-08 1.580E-07 2.973E-07 2.979E-07 2.234E-07 1.972E-07 1.866E-07 1.947E-07 3.038E-07 3.970E-07 4.294E-07 2.095E-07 1.658E-07 9.121E-08 4.960E-08 2.846E-08 5.184E-08 9.714E-08 1.832E-07 1.827E-07 1.364E-07 1.204E-07 1.140E-07 1.184E-07 1.849E-07 2.435E-07 2.653E-07 1.494E-07 1.186E-07 6.504E-08 3.536E-08 2.021E-08 3.662E-08 6.883E-08 1.301E-07 1.292E-07 9.622E-08 8.490E-08 8.045E-08 8.330E-08 1.301E-07 1.722E-07 1.886E-07 1.150E-07 9.147E-08 5.005E-08 2.721E-08 1.550E-08 2.799E-08 5.272E-08 9.973E-08 9.881E-08 7.345E-08 6.478E-08 6.141E-08 6.343E-08 9.912E-08 1.317E-07 1.448E-07 9.290E-08 7.402E-08 4.042E-08 2.197E-08 1.249E-08 2.249E-08 4.242E-08 8.032E-08 7.940E-08 5.895E-08 5.196E-08 4.929E-08 5.080E-08 7.942E-08 1.059E-07 1.168E-07 7.760E-08 6.191E-08 3.376E-08 1.835E-08 1.041E-08 1.870E-08 3.532E-08 6.692E-08 6.604E-08 4.898E-08 4.316E-08 4.095E-08 4.214E-08 6.590E-08 8.809E-08 9.740E-08 6.641E-08 5.304E-08 2.889E-0 8 1.570E-08 8.897E-09 1.595E-08 3.014E-08 5.714E-08 5.630E-08 4.172E-08 3.675E-08 3.489E-08 3.584E-08 5.607E-08 7.511E-08 8.325E-08 5.789E-08 4.628E-08 2.518E-08 1.368E-08 7.744E-09 1.386E-08 2.621E-08 4.927E-08 4.892E-08 3.623E-08 3.189E-08 3.029E-08 3.108E-08 4.863E-08 6.527E-08 7.248E-08 5.121E-08 4.097E-08 2.227E-08 1.209E-08 6.841E-09 1.222E-08 2.314E-08 4.390E-08 4.314E-08 3.193E-08 2.810E-08 2.670E-08 2.736E-08 4.282E-08 5.757E-08 6.405E-08 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-113 Table 2.3-38b CGS Calculation, Terrain Features, Desert Sigmas

CGS TURBINE AND RADWASTE BLDGS NO DECAY, UNDEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT

SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 2.782E-05 2.117E-05 1.211E-05

6.468E-06

4.034E-06

7.812E-06

1.386E-05

2.549E-05

2.640E-05

2.061E-05

1.800E-05

1.715E-05

1.821E-05

2.834E-05

3.509E-05

3.628E-05 7.806E-06 6.000E-06 3.404E-06

1.831E-06

1.113E-06

2.127E-06

3.830E-06

7.081E-06

7.275E-06

5.617E-06

4.929E-06

4.677E-06

4.961E-06

7.730E-06

9.967E-06

1.013E-05 2.832E-06 2.195E-06 1.236E-06

6.680E-07

3.982E-07

7.518E-07

1.370E-06

2.548E-06

2.599E-06

1.986E-06

1.749E-06

1.656E-06

1.751E-06

2.730E-06

3.466E-06

3.656E-06 1.567E-06 1.220E-06 6.834E-07

3.702E-07

2.186E-07

4.096E-07

7.517E-07

1.402E-06

1.424E-06

1.082E-06

9.543E-07

9.030E-07

9.521E-07

1.484E-06

1.899E-06

2.015E-06 1.037E-06 8.099E-07 4.521E-07

2.451E-07

1.439E-07

2.682E-07

4.944E-07

9.242E-07

9.356E-07

7.085E-07

6.251E-07

5.914E-07

6.221E-07

9.699E-07

1.247E-06

1.330E-06 5.081E-07 3.989E-07 2.214E-07

1.202E-07

6.994E-08

1.292E-07

2.397E-07

4.498E-07

4.528E-07

3.410E-07

3.009E-07

2.848E-07

2.982E-07

4.651E-07

6.035E-07

6.485E-07 2.113E-07 1.671E-07 9.199E-08

5.001E-08

2.873E-08

5.239E-08

9.809E-08

1.849E-07

1.845E-07

1.379E-07

1.217E-07

1.152E-07

1.198E-07

1.870E-07

2.459E-07

2.677E-07 1.153E-07 9.172E-08 5.019E-08

2.729E-08

1.555E-08

2.809E-08

5.290E-08

1.001E-07

9.915E-08

7.372E-08

6.502E-08

6.164E-08

6.368E-08

9.951E-08

1.322E-07

1.453E-07 7.771E-08 6.199E-08 3.381E-08

1.837E-08

1.043E-08

1.873E-08

3.538E-08

6.703E-08

6.615E-08

4.906E-08

4.323E-08

4.103E-08

4.221E-08

6.602E-08

8.823E-08

9.755E-08 5.794E-08 4.631E-08 2.520E-08

1.369E-08

7.751E-09

1.387E-08

2.624E-08

4.976E-08

4.897E-08

3.626E-08

3.193E-08

3.032E-08

3.111E-08

4.868E-08

6.534E-08

7.255E-08

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-114 Table 2.3-38c CGS Calculation, Terrain Features, Desert Sigmas CGS TURBINE AND RADWASTE BLDGS - SPECIFIC POINTS OF INTEREST RELEASE ID TYPE OF LOCATION DIRECTION FROM SITE DISTANCE X/Q X/Q X/Q D/Q (MILES) (METERS) (SEC/CUB.METER) NO DECAY UNDEPLETED (SEC/CUB.METER) 2.260 DAY DECAY UNDEPLETED (SEC/CUB.METER) 8.000 DAY DECAY DEPLETED (PER SQ.METER)

A A A

A A

A A

A A

A A

A A

A A

A A

A A

A A

A A

A A A A A

A A

A A

A A

A A

A A

A A PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

EAB (1950 M) EAB (1950 M) EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

SE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE SE SE ESE ENE NNE ESE SE SE NE S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 6.40 3.90 4.00 4.10 4.20 4.30 4.40 4.50 3.00 3.10 3.21 3.30 3.40 3.50 3.60 15.00 4.80 4.20 4.10 4.30 0.10 9.59 8.29 4.30 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 10298. 6275.

6436.

6597.

6758.

6919.

7080.

7241.

4827.

4988.

5159.

5310.

5471.

5632.

5792. 24135. 7723.

6758.

6597.

6918. 160. 15437.

13346. 6919.

1950. 1950. 1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950. 7.2E-07 1.2E-06 1.2E-06 1.1E-06 1.1E-06 1.0E-06 1.0E-06 9.6E-07 1.9E-06 1.8E-06 1.7E-06 1.6E-06 1.5E-06 1.5E-06 1.4E-06 2.4E-07 1.1E-06 1.1E-06 6.9E-07 7.6E-07 4.1E-04 4.2E-07 5.1E-07 6.7E-07 1.1E-05 8.4E-06 4.8E-06 2.6E-06 1.6E-06 3.0E-06 5.4E-06 9.9E-06 1.0E-05 7.9E-06 6.9E-06 6.6E-06 7.0E-06 1.1E-05 1.4E-05 1.4E-05 6.8E-07 1.2E-06 1.1E-06 1.1E-06 1.0E-06 1.0E-06 9.6E-07 9.2E-07 1.9E-06 1.8E-06 1.7E-06 1.6E-06 1.5E-06 1.4E-06 1.4E-06 2.1E-07 1.1E-06 1.0E-06 6.5E-07 7.3E-07 4.1E-04 3.9E-07 4.7E-07 6.3E-07 1.1E-05 8.3E-06 4.7E-06 2.5E-06 1.5E-06 2.9E-06 5.3E-06 9.8E-06 1.0E-05 7.8E-06 6.8E-06 6.5E-06 6.9E-06 1.1E-05 1.3E-05 1.4E-05 5.1E-07 9.3E-07 8.9E-07 8.5E-07 8.1E-07 7.8E-07 7.5E-07 7.2E-07 1.5E-06 1.4E-06 1.3E-06 1.3E-06 1.2E-06 1.1E-06 1.1E-06 1.5E-07 8.3E-07 8.1E-07 5.2E-07 5.7E-07 4.0E-04 2.8E-07 3.5E-07 5.0E-07 9.4E-06 7.2E-06 4.1E-06 2.2E-06 1.3E-06 2.6E-06 4.6E-06 8.5E-06 8.8E-06 6.8E-06 5.9E-06 5.6E-06 6.0E-06 9.3E-06 1.2E-05 1.2E-05 2.6E-10 7.0E-10 6.6E-10 6.3E-10 6.0E-10 5.7E-10 5.4E-10 5.1E-10 1.3E-09 1.2E-09 1.1E-09 1.0E-09 9.6E-10 9.0E-10 8.4E-10 6.0E-11 4.9E-10 6.0E-10 3.7E-10 5.1E-10 8.4E-07 1.3E-10 1.7E-10 3.7E-10 7.9E-09 5.5E-09 3.5E-09 1.4E-09 1.3E-09 2.5E-09 5.2E-09 1.1E-08 1.3E-08 1.0E-08 7.3E-09 6.7E-09 7.0E-09 1.1E-08 1.2E-08 1.1E-08 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-115 Table 2.3-38d CGS Calculation, Terrain Features, Desert Sigmas CGS REACTOR BLDG NO DECAY, UNDEPLETED CORRECTED USING STANDARD OPEN TERRAIN FACTORS ANNUAL AVERAGE CHI/Q (SEC/METER CUBED)

DISTANCE IN MILES FROM THE SITE SECTOR 0.250 0.500 0.750 1.000 1.500 2.000 2.500 3.000 3.500 4.000 4.500 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 3.983E-07 3.015E-07 2.241E-07 9.654E-08 3.448E-08 7.411E-08 2.513E-07 7.930E-07 1.124E-06 1.049E-06 6.008E-07 4.892E-07 5.015E-07 9.120E-07 6.666E-07 4.473E-07 5.147E-07 3.460E-07 2.226E-07 7.983E-08 7.360E-08 1.508E-07 3.020E-07 7.233E-07 8.905E-07 7.337E-07 4.702E-07 5.305E-07 5.031E-07 7.571E-07 6.348E-07 5.728E-07 4.459E-07 2.917E-07 1.853E-07 7.694E-08 7.682E-08 1.476E-07 2.641E-07 5.856E-07 6.762E-07 5.603E-07 3.787E-07 4.710E-07 4.747E-07 6.975E-07 5.775E-07 5.291E-07 2.855E-07 1.836E-07 1.175E-07 5.258E-08 5.380E-08 1.017E-07 1.670E-07 3.550E-07 4.031E-07 3.353E-07 2.340E-07 3.229E-07 3.390E-07 4.987E-07 3.846E-07 3.498E-07 1.417E-07 8.972E-08 5.986E-08 2.801E-08 2.883E-08 5.532E-08 8.155E-08 1.649E-07 1.877E-07 1.577E-07 1.123E-07 2.809E-07 3.184E-07 4.778E-07 2.045E-07 1.780E-07 8.546E-08 5.399E-08 3.986E-08 1.857E-08 1.888E-08 3.697E-08 4.940E-08 9.701E-08 1.113E-07 9.422E-08 6.758E-08 3.180E-07 3.702E-07 5.635E-07 1.394E-07 1.089E-07 6.007E-08 3.809E-08 3.227E-08 1.473E-08 1.476E-08 2.952E-08 3.530E-08 6.783E-08 7.839E-08 6.677E-08 4.829E-08 3.667E-07 4.232E-07 6.623E-07 1.150E-07 7.782E-08 4.723E-08 3.012E-08 2.901E-08 1.300E-08 1.283E-08 2.619E-08 2.826E-08 5.335E-08 6.206E-08 5.308E-08 3.883E-08 5.692E-07 6.426E-07 1.026E-06 1.046E-07 6.216E-08 3.955E-08 2.537E-08 2.518E-08 1.120E-08 1.088E-08 2.256E-08 2.409E-08 4.477E-08 5.232E-08 4.498E-08 4.067E-08 1.128E-06 1.325E-06 1.950E-06 2.207E-07 5.287E-08 3.450E-08 2.223E-08 2.229E-08 9.892E-09 9.460E-09 1.990E-08 2.133E-08 3.913E-08 4.585E-08 3.965E-08 4.490E-08 1.019E-06 1.054E-06 1.650E-06 7.638E-07 4.674E-08 3.085E-08 1.996E-08 2.000E-08 8.868E-09 8.362E-09 1.781E-08 1.931E-08 4.183E-08 4.113E-08 4.297E-08 8.482E-08 8.387E-07 8.662E-07 1.356E-06 6.314E-07 4.227E-08 ANNUAL AVERAGE CHI/Q (SEC/METER CUBED) DISTANCE IN MILES FROM THE SITE SECTOR 5.000 7.500 10.000 15.000 20.000 25.000 30.000 35.000 40.000 45.000 50.000

S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 2.814E-0 8 1.825E-08 1.814E-08 8.058E-09 7.499E-09 1.613E-08 1.778E-08 4.509E-08 3.754E-08 4.676E-08 2.656E-07 7.075E-07 7.296E-07 1.143E-06 5.345E-07 3.896E-08 2.023E-08 1.646E-08 1.472E-08 6.624E-09 4.948E-09 1.203E-08 1.734E-08 2.990E-08 4.858E-08 4.163E-07 4.104E-07 3.901E-07 4.000E-07 6.275E-07 2.974E-07 3.642E-08 1.577E-08 1.027E-08 6.195E-08 2.853E-08 3.627E-09 1.269E-08 1.283E-08 2.199E-08 4.276E-07 3.178E-07 2.789E-07 2.653E-07 2.710E-07 1.652E-07 2.034E-07 2.261E-08 3.200E-07 2.576E-07 1.393E-07 7.526E-08 4.242E-08 7.524E-08 1.423E-07 2.697E-07 2.625E-07 1.945E-07 1.705E-07 1.623E-07 1.650E-07 2.594E-07 3.525E-07 3.975E-07 2.287E-07 1.846E-07 9.960E-08 5.376E-08 3.020E-08 5.333E-08 1.011E-07 1.919E-07 1.860E-07 1.376E-07 1.205E-07 1.148E-07 1.163E-07 1.830E-07 2.499E-07 2.834E-07 1.764E-07 1.427E-07 7.682E-08 4.145E-08 2.323E-08 4.089E-08 7.760E-08 1.474E-07 1.425E-07 1.053E-07 9.214E-08 8.786E-08 8.872E-08 1.398E-07 1.916E-07 2.181E-07 1.428E-07 1.156E-07 6.218E-0 8 3.353E-08 1.876E-08 3.294E-08 6.258E-08 1.190E-07 1.148E-07 8.473E-08 7.408E-08 7.067E-08 7.121E-08 1.123E-07 1.544E-07 1.762E-07 1.195E-07 9.686E-08 5.202E-08 2.804E-08 1.567E-08 2.746E-08 5.221E-08 9.930E-08 9.564E-08 7.055E-08 6.164E-08 5.884E-08 5.918E-08 9.334E-08 1.287E-07 1.473E-07 1.024E-07 8.309E-08 4.458E-08 2.402E-08 1.341E-08 2.346E-08 4.463E-08 8.493E-08 8.168E-08 6.022E-08 5.259E-08 5.021E-08 5.043E-08 7.958E-08 1.099E-07 1.261E-07 8.938E-08 7.259E-08 3.891E-08 2.096E-08 1.169E-08 2.042E-08 3.888E-08 7.401E-08 7.108E-08 5.238E-08 4.572E-08 4.367E-08 4.380E-08 6.915E-08 9.568E-08 1.099E-07 7.915E-08 6.433E-08 3.446E-08 1.856E-08 1.034E-08 1.805E-08 3.437E-08 6.544E-08 6.278E-08 4.625E-08 4.035E-08 3.855E-08 3.862E-08 6.099E-08 8.452E-08 9.728E-08

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-116 Table 2.3-38e CGS Calculation, Terrain Features, Desert Sigmas

CGS REACTOR BLDG NO DECAY, UNDEPLETED CHI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECTION FROM SITE .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 3.899E-07

2.557E-07

1.635E-07 6.676E-08 6.588E-08

1.279E-07

2.294E-07

5.137E-07

6.024E-07

4.988E-07

3.347E-07

4.184E-07

4.207E-07

6.224E-07

5.045E-07

4.591E-07 1.486E-07

9.471E-08

6.378E-08 2.927E-08 2.996E-08

5.746E-08

8.625E-08

1.770E-07

2.016E-07

1.690E-07

1.195E-07

3.067E-07

3.460E-07

5.205E-07

2.156E-07

1.855E-07 6.171E-08

3.914E-08

3.299E-08 1.506E-08 1.509E-08

3.018E-08

3.624E-08

6.982E-08

8.063E-08

6.861E-08

4.965E-08

4.347E-07

4.968E-07

7.813E-07

1.174E-07

7.985E-08 3.982E-08

2.553E-08

2.517E-08 1.122E-08 1.090E-08

2.258E-08

2.423E-08

4.507E-08

5.264E-08

4.526E-08

4.175E-08

9.267E-07

1.027E-06

1.572E-06

3.944E-07

5.319E-08 3.093E-08

2.000E-08

1.999E-08 8.872E-09 8.368E-09

1.781E-08

1.934E-08

4.224E-08

4.120E-08

4.339E-08

1.400E-07

8.436E-07

8.714E-07

1.364E-06

6.347E-07

4.237E-08 2.000E-08

1.411E-08

3.647E-08 1.668E-08 4.928E-09

1.324E-08

1.543E-08

2.976E-08

2.146E-07

2.904E-07

3.198E-07

4.052E-07

4.159E-07

5.365E-07

3.083E-07

3.085E-08 2.118E-07

1.702E-07

1.045E-07 5.532E-08 2.837E-08

5.160E-08

9.519E-08

1.801E-07

2.652E-07

1.966E-07

1.723E-07

1.641E-07

1.669E-07

2.045E-07

2.738E-07

2.635E-07 1.769E-07

1.431E-07

7.704E-08 4.156E-08 2.330E-08

4.103E-08

7.785E-08

1.479E-07

1.430E-07

1.057E-07

9.247E-08

8.817E-08

8.906E-08

1.403E-07

1.923E-07 2.188E-07 1.196E-07

9.698E-08

5.209E-08 2.808E-08 1.569E-08

2.750E-08

5.228E-08

9.945E-08

9.579E-08

7.066E-08

6.174E-08

5.893E-08

5.928E-08

9.350E-08

1.289E-07

1.475E-07 8.944E-08

7.264E-08

3.894E-08 2.098E-08 1.170E-08

2.044E-08

3.891E-08

7.407E-08

7.115E-08

5.243E-08

4.576E-08

4.371E-08

4.385E-08

6.922E-08

9.576E-08

1.100E-07

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-117 Table 2.3-38f CGS Calculation, Terrain Features, Desert Sigmas CGS REACTOR BLDG, SPECIFIC POINTS OF INTEREST RELEASE ID TYPE OF LOCATION DIRECTION FROM SITE DISTANCE X/Q X/Q X/Q D/Q (MILES) (METERS) (SEC/CUB.METER)

NO DECAY UNDEPLETED (SEC/CUB.METER) 2.260 DAY DECAY UNDEPLETED (SEC/CUB.METER) 8.000 DAY DECAY DEPLETED (PER SQ.METER)

B B B

B B

B B

B B

B B

B B

B B

B B

B B

B B

B B

B B B B B

B B

B B

B B

B B

B B

B B PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE PROTECTED ARE LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

LPZ (4828)

EAB (1950 M) EAB (1950 M) EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

EAB (1950 M)

SE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE ESE SE SE ESE ENE NNE ESE SE SE NE S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE SE SSE 6.40 3.90 4.00 4.10 4.20 4.30 4.40 4.50 3.00 3.10 3.21 3.30 3.40 3.50 3.60 15.00 4.80 4.20 4.10 4.30 0.10 9.59 8.29 4.30 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 1.21 10298. 6275.

6436.

6597.

6758.

6919.

7080.

7241.

4827.

4988.

5159.

5310.

5471.

5632.

5792. 24135. 7723.

6758.

6597.

6918. 160. 15437.

13346. 6919.

1950. 1950. 1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950.

1950. 3.7E-07 1.7E-06 1.7E-06 1.6E-06 1.5E-06 1.5E-06 1.4E-06 1.4E-06 1.0E-06 1.2E-06 1.5E-06 1.8E-06 1.9E-06 1.8E-06 1.8E-06 3.5E-07 5.7E-07 1.5E-06 9.8E-07 4.1E-08 3.1E-06 2.1E-07 2.6E-07 6.8E-08 1.9E-07 1.2E-07 7.9E-08 3.8E-08 3.8E-08 7.2E-08 1.1E-07 2.3E-07 2.6E-07 2.2E-07 1.6E-07 2.7E-07 2.9E-07 4.4E-07 2.7E-07 2.4E-07 3.5 E-07 1.6E-06 1.6E-06 1.5E-06 1.5E-06 1.4E-06 1.3E-06 1.3E-06 9.9E-07 1.2E-06 1.5E-06 1.8E-06 1.8E-06 1.8E-06 1.7E-06 3.1E-07 5.5E-07 1.5E-06 9.3E-07 4.1E-08 3.1E-06 2.0E-07 2.5E-07 6.6E-08 1.9E-07 1.2E-07 7.9E-08 3.8E-08 3.7E-08 7.1E-08 1.1E-07 2.3E-07 2.6E-07 2.2E-07 1.6E-07 2.7E-07 2.9E-07 4.4E-07 2.7E-07 2.4E-07 3.4E-07 1.3E-06 1.3E-06 1.2E-06 1.2E-06 1.1E-06 1.1E-06 1.0E-06 8.1E-07 9.6E-07 1.2E-06 1.5E-06 1.5E-06 1.4E-06 1.4E-06 2.9E-07 5.5E-07 1.2E-06 7.7E-07 3.9E-08 3.1E-06 1.9E-07 2.4E-07 6.6E-08 1.8E-07 1.2E-07 7.6E-08 3.6E-08 3.7E-08 7.0E-08 1.1E-07 2.2E-07 2.5E-07 2.1E-07 1.5E-07 2.6E-07 2.9E-07 4.3E-07 2.6E-07 2.3E-07 3.2E-10 7.0E-10 6.6E-10 6.4E-10 6.0E-10 5.7E-10 5.5E-10 5.2E-10 1.2E-09 1.2E-09 1.1E-09 1.0E-09 9.4E-10 8.9E-10 8.3E-10 7.6E-11 5.9E-10 6.0E-10 3.8E-10 1.8E-10 4.0E-08 1.6E-10 2.0E-10 1.3E-10 1.6E-09 1.0E-09 5.7E-10 1.9E-10 2.4E-10 4.8E.10 9.0E-10 2.0E-09 2.8E-09 2.5E-09 1.8E-09 1.8E-09 1.5E-09 2.4E-09 2.0E-09 1.9E-09 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-118

TABLE 39 (a through f)

DELETED C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-119 Table 2.3-40 Frequency of Wind Resuspension Periods at Hanf ord (1953-1970)

Total Dust Hours 476 Total Dust Days 142 Number of Dust Storms 150 Average Dust Hr/Yr.

26.4 Average Dust Days/Yr.

7.9 Average

Dust Storms Per Year

8.3 Range

in Duration of Du st Storms (hr.)

1-16 Average Duration of Du st Storms (hr.)

3.2 Average

Dust Storm Concentration (from Table 2) mg/m 3 6.77

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-120 Table 2.3-41 Dust Concentration Dependency on Wind Speed and Direction at Hanford 1953-1970 Predicted Concentration From Visibility, mg/m 3 WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 WIND SPEED CLASS (MPH) 55-63 64-UP OVERALL AVERAGE 25-31 32-38 39-46 47-54 SE SSE S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE .00 .00 .00

.00 .00 .00 .00 .00

.00 .00 .00

.00 .00 .00 .00 .00 .00 .00 .00

.00 .00 .00 1.74 .00 3.29 2.02 3.29 1.74 4.38 .00

.00 2.71 .00

.00 7.83 .00 2.71 1.74 1.74 3.49 1.88 2.60 2.92 3.38 4.60 .00 3.29 .00 .00 2.71 1.62 2.48 6.34 1.81 1.83 2.64 2.58 2.58 3.50 3.41 3.38 3.05 2.44 .00 .00 1.38 1.38 1.62 3.54 4.96 2.89 1.77 1.50 4.80 5.06 6.08 4.54 2.19 3.60 .00 .00 1.25 5.70 2.21 2.75 4.13 5.37 1.99 1.98 .00 12.99 7.04 2.81 .00 2.71 .00 .00

.00 15.92 3.86 8.83 12.95 2.71 3.29 2.23 .00

.00 7.83 .00

.00 .00 .00 .00 .00 .00 4.13 13.87 48.31 .00 4.13 .00 .00 .00

.00 .00 .00 .00 .00 .00 .00 .00

.00 .00 .00 .00 .00

.00 .00 .00

.00 .00 .00 .00 .00 .00 .00 .00

.00 988.88* .00 .00 .00

.00 .00 .00

.00 .00 .00 .00 .00 .00 .00 .00

.00 .00 .00 .00 .00

.00 .00 .00

.00 .00 .00 .00 .00 .00 1.78 7.17 2.95 19.40 7.67 3.54 2.39 2.08 2.77 3.81 4.77 3.84 2.48 2.78 2.71 OVERALL** AVERAGE .00 2.84 3.00 3.15 4.15 3.71 8.57 22.22 .00 988.88 .00 6.77

.00 NO DATA

• VISIBILITY 0 TO 1/16 MILE DUE TO ONE-HOUR DUSTSTORM

• • WEIGHTED AVERAGE BASED ON TABLE 2.3-42 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.3-121 Table 2.3-42 Hours Satisfying Dust Storm Criteria at Hanford (1953-1970) Hours With (1) Visibility 7 Mile and Dust Reported or (2) Visibility 7 to 14 Miles, Windspeed 5.8 M/Sec: RH 70% Dust Assumed WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 WIND SPEED CLASS (MPH) 55-63 64-UP TOTAL HOURS 25-31 32-38 39-46 47-54 SE SSE S SSW SW WSW W WNW NW NNW N NNE NE ENE E ESE 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 1 0 1 2 1 1 2 0 0 1 0 0

1 0 1 1 1 5 6 8 12 3 3

0 1 0 0 1

3 4 3 7 3 5 4 6 34 31 19 3 6 0 0 1

1 3 13 17 5 11 3 2 10 23 15 6 2 0 0 1

3 11 24 39 7 6 5 2 1 7 5 0 1 0 0 0

3 13 26 13 1 1 2 0 0 1 0 0 0 0 0 0

0 2 6 4 0 1 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 3 11 33 74 81 18 29 21 18 58 66 44 9 10 1 TOTAL HOURS 0 9 42 129 112 110 60 13 0 1 0 476

QQ QQ QQ QQ

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 2.4-1 2.4 HYDROLOGY ENGINEERING

The italicized information is historical and was provided to support the application for an operating license.

2.4.1 HYDROLOGIC

DESCRIPTION

2.4.1.1 Site and Facilities Columbia Generating Station (C GS) is located in the Hanfor d Site within Benton County, Washington, approximately 3 mile s west of the Columbia River at river mile (RM) 352, 10 miles north of Richland and 45 miles downstream from Grant County PUD Pr iest Rapids Dam. The site coordinates are appr oximately 46° 28' Nort h Latitude and 119° 20' West Longitude.

The Columbia River is the predominant hydrologic feature of th e area and provides principal drainage for the surrounding area.

The riverbed is clearly mark ed in the terrain and at the

proximity of the site the river flows between high banks. The Columb ia River approximate riverbed elevation is 328 ft above mean sea le vel (msl); the ground elev ation at the site is approximately 440 ft. Another hydr ologic feature of the area is th e Yakima River, which at its closest approach flows within 8 miles of the plant site. The river system is shown in the hydrographic map, Figure 2.4-1. Figures 2.1-1 and 2.1-2 show the major hydrologic features of the area.

All Seismic Category I structures are located above maximum pos tulated flood elevations. For flood elevations refer to Sections 2.4.3 and 2.4.4.

Water for cooling tower makeup water and other plant requireme nts is withdrawn from the Columbia River. The intake system is designed for a maximum capacity of 25,000 gpm (55.7 cfs). The non-safety-relat ed makeup water intake system is approximately 3 miles east of the plant and is made up of two offshore perforated pipe inle ts, two lead-in pipes, and pump house structure.

A topographic map and contour map of the region surrounding the site are shown in

Figures 2.4-2 and 2.4-3. The natural drainage features of the surrounding area have not been changed by the cons truction of CGS.

2.4.1.2 Hydrosphere

The Columbia River, the largest river flowing into the Pacific Ocean from North America, is one of this world's greatest sources of hydroelectric power. Its annual discharge of 18,000,000 acre ft (1 acre-ft = 43,560 ft

3) is exceeded in the No rth American continent only by the Mississippi, Mackenzi e and St. Lawrence Rivers.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.4-2 The Columbia River drains an area of approximately 258,000 square m iles, lying to the west of the Continental Divide in the northwestern part of the U.S. (85%) and southwestern part of Canada (15%). Major tributaries are th e Kootenay, Snake, Pend Oreille, Spokane, Okanogan, Yakima, and Willamette Rivers.

In determining the Standard Pr oject Flood (SPF) the drainage area was divided into subbasins.

These subbasins can be grouped into six general areas w ith similar hydrometeorological characteristics.

The six areas are: (1) upper Columbia, which includes the drainage of the area in Canada and the northern part of the United States above Ch ief Joseph Dam; (2) Middle Columbia, which includes the area between Pasco and Chief Joseph Dam; (3) Upper and Middle Snake River; (4) Lower Snake River, the area between Weiser and Ice Harbor Dam; (5) Lower Columbia, including the area between Bonneville Dam and Pasco; (6) the Columb ia below Bonneville Dam, including the Willamette River.

The river basin has five outstanding physical features: the Rocky Mountain System, the Columbia Plateau, the Colu mbia River Gorge, the Cascade Range and Puget Trough.

The Rocky Mountain System is the major range with elevations from 2000 to over 12,000 ft.

There are permanent glaciers and extensive snow fields at highe r elevations and deep valleys that provide the principal drai nage for the head-waters of the Columbia, Kootenay and other rivers.

The Columbia Plateau is a great, generally treeless, semiarid plateau covering over 100,000 square miles in the centr al portion of the basin. This plateau is in an area between the Cascade Range and the Rocky Mountains. The pl ateau was formed by successive flows of lava and filled to a general thickness of approximately 4000 ft. The Columbia River flows 1214 miles from its source in Columbia Lake (el. 2700 ft) in British Columbia, near the crest of the Rockies, to the Pacific Ocean at Astori a, Oregon. It sweeps around the north and northwesterly sides of the Colu mbia Plateau to central Washi ngton to be joined by the Snake River. The Columbia River flows directly across the axis of the Cascades in a narrow gorge to the Pacific.

The Columbia Gorge is the gat eway from the Pacific Ocean to the intermountain Columbia Plateau. Tide flows 140 miles up-river. For most of its length the river flows in deep valleys

and canyons.

High flows occur in late spring and early su mmer with melting of s now on the mountainous watershed. Low flows occur in autumn and winter.

The Columbia River has been regulated by dams and reservoirs over the past 35 years. A large portion of the main stream and major tributaries is devel oped to meet various 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 2.4-3 functional requirements, such as flood control, hydroelectric power, irrigation, municipal and industrial supply, etc. The regulation of Columbia River floods is accomplished by use of reservoir storage space pr ovided primarily for irrigation or for hydroelectric power utilization.

The volume of usable reservoir storage space is on the order of 20% to 25%.

There are seven dams upstream and four dams downs tream of the site on the main stream of the Columbia River within the U.S. These dams are listed in Table 2.4-1. The Columbia River flow in the reach of CGS is controlled by regul ation of the upstream re servoir projects, which have a total usable storage capacity of approximately 35 million ac re-ft. Some control of flow in the immediate vicinity of th e site is by regulation of th e nearest upstream hydroelectric projects, Priest Rapids Dam, at RM 397, containing about 45, 000 acre-ft of active storage, and Wanapum Dam, at RM 415, containi ng about 161,000 ac re-ft of active storage. Some minimal effect on the river flow in the vicinity of the site is caused by McNary Dam, at RM 292, approximately 60 RM downs tream from the site area.

Flows in the Columbia River during the summer, fall, and wi nter vary from a low of 36,000 ft 3/sec to as much as 160,000 ft 3/sec. During spring runo ff high flows ranging from 250,000 ft 3/sec to 450,000 ft 3/sec have been recorded.

The average annual flow is 120,000 ft 3/sec; during low flow periods flo ws may average about 60,000 ft 3/sec (see Figure 2.4

-4).

The Grand Coulee and Bonneville dams were put into operation prior to World War II and several dams were built after the war. The four downstream dams include large locks to permit the passage of river ve ssels. Several of the dams provide emergency floodwater storage. Grand Coulee, the largest and most co mplex of the dams, augments the low winter flows for the entire system from its 9,402,00 0 acre-ft of available storage (of which approximately 5,100,000 acre-ft is active storage) and also pumps water to the Columbia River Irrigation Project.

The river channel near the CGS site varies betwee n 400 and 600 yards in width for low water and normal high water level, resp ectively. The depth varies from about 25 ft to 45 ft for normal high water and flood high wate r levels, respec tively. Velocities va ry from 3 ft/sec to over 11 ft/sec depending on section and flow. Average water temperature is 51°F.

Temperatures may reach a low of 32°F and a high of 68°F. (See Table 2.4-2 and Figure 2.4

-4.)

A list of water usage downstream of CGS, obtained from records of the Department of Ecology, State of Washington, for water rights as of February 1980, is presented in Table 2.4-3. The closest municipal surface water user is the C ity of Richland with an intake approximately 12 miles downstream.

The location of local groundw ater users is discussed in Section 2.4.13.2.

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 2.4-4 2.4.2 FLOODS

2.4.2.1 Flood History

Floods in the Columbia Ri ver Basin are grouped as:

a. The interior basin east of the Ca scades, caused by melting snowpack and occurring from May through June;

b. The Willamette and other basins, west of the Cascades, caused by direct runoff from intense winter rain occa sionally augmented by snowmelt.

There is some overlapping effect within these two groupings. At certain elevations, basins in the interior Columbia drainage area occasionally have signific ant flood flows resulting from winter rain or snowmelt. Thes e are local floods and do not usually contribute sufficient flow to cause flooding of the main Columbia River. Major floods in the Columbia River Basin result from rapid spring melting of the snowpack over a wide area, generally augmented by rain or by above-normal precipitation in May, accompanied by a major chinook wind which causes rapid area temperature rise.

The annual spring snowmelt flood of the main interior basin is characterized by relatively un iform distribution over the bas in. The snowfall and individual snow storms may vary, but the in tegration of all storms over the winter period smoothes the irregularities, with the result that the distri bution of the flood runoff is reasonably constant from year to year.

The maximum historical flood of record is th at of June 7, 1894, which resulted from a combination of hydrometeorologic conditions, in cluding heavy snowpack and rapid melt plus rainfall. The peak disc harge at CGS was 740,000 ft 3/sec for the Columbia River, as estimated from high water marks at Wena tchee, Washington (Reference 2.4-1). The largest recent flood, occurring in 1948, had an observed peak discharge of 690,000 ft 3/sec at Hanford. These floods were spring floods resulting from the melt of a large snowpack combined with the spring rains (Reference 2.4-2). Water surface profiles for the Columbia River in the vicinity of the site as derived by the Cor p s of Engineers (Refe r ence 2.4-2) are given in Figure 2.4

-5.

The plant site is located approxi mately three miles west of the Columbia River at RM 352 with reactor floor elevation of 441 ft msl, which is 68 ft above the water level estimated for the largest historical flood (approxi mately 373 ft msl). There is no record of flooding in the immediate site area.

2.4.2.2 Flood Design Considerations

Flood protection of safety-related components is based on the highest calculated flood water level including wave effects, resulting from intense local precipitation.

Several different probable maximum events were considered, including the Corps of Engineers design-project

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-069 2.4-5 flood considered to be "the mo st severe reasonably possible.

" Wave action caused by storm winds, the effects of failure of upstream dam surge floodi ng, and ice flooding were also considered.

The results of these analyses (described in Section 2.4.3) indicate that the CGS site is safe from floods and that no flood pr otection measures are required.

The Hydrogen Storage and Supply Facility located 0.6 miles south-southeast of the plant is subject to flooding due to the PMP flood discussed in Section 2.4.3.1. Equipment storing liqui d and gaseous hydrogen has been analyzed for the effe cts of this flood (Section 2.4.2.3). As discussed in sections that follow, plant safety-relat ed structures are located above high water elevations associated with Columbia River flooding (Sections 2.4.2.1 and 2.4.3), intense local precipitation (Sections 2.4.3.5 and 2.4.3.6), and upriver dam failures (Section 2.4.4).

2.4.2.3 Effects of Local Intense Precipitation

Intense local summer thunder st orms can produce short duration rains which have the potential for causing serious flood. Winter precipitation may occur as rain or snow and would be less intense than the worst summer thunderstorm. The probable maximum precipitation (PMP)

event for the CGS site has been determined using the methodology deve loped by the U.S. Weather Bureau and reported in Hydrometeorological Report No. 43, "Probable Maximum Precipitation, Northwest States" (Reference 2.4-3).

The plant area slopes easterly to a broad channe l which is adequate to store and drain the PMP. Construction contours of the site are shown in Figure 2.4-28. The reactor building and the spray ponds are located at elevations that are safe from the effect of any flood resulting from the maximum precipitation event.

Winter precipitation may occur as rain or snow. The winter season s nowfall has ranged from less than 0.5 in. to a maximum of 12 in. in December 1964. There is no ice accumulation at the site.

To accommodate surface drainage during severe climatic conditions such as rainfall and rapid snow melts, a system of catch basins and dry we lls is provided with inlet elevations a minimum of 6 in. lower than the nearest road and a minimum of 12 in. lower than the finished floor elevation of the nearest building(s).

Runoff from the PMP event is accommodated by designing the roadways such that the high point of the road is 6 in. to 1 ft below the fini shed floor elevation of the adjacent safety-related building(s). Runoff from this event is from the northwest to the s outheast across the site plateau to the low area southeast of the plant si te. The general plant s ite is nominally 9 ft above the maximum calculated water surface elevation resulting from the postulated PMP (Section 2.4.3.3). Therefore, the si te grading precludes the potentia l flooding of safety-related structures. The Hydrogen Storage and Supply Faci lity is subject to th e PMP event flooding.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-13-050 2.4-6 The elevation of the facility is 420 ft ms l and the PMP flood level is 431.1 ft msl (Section 2.4.3.5). See Section 10.4.10 for more discussion.

Roofs of buildings are designed to take, with adequate drainage, any instantaneous or local intense precipitation. Discharge from roof drai ns is carried by means of a storm sewer system to a manhole located southeast of the reactor building. From that point a pipeline with a northeast alignment transf ers the discharge to lin ed evaporation ponds about 1500 ft away from the plant site.

The roofs of safety-related buildings (diesel generator building, radw aste/control building, standby service water pump house) are concrete beam and slab construction except the high roof of the reactor building, which is metal de ck on steel framing.

The minimum roof slope for all structures is 1/8 in. per ft for adequate drainage and the roof areas are encompassed by curbs or parapet walls up to 3 ft 6 in. high. Roof plans, including details of roof drains and overflow scuppers, are provided in Figure 2.4-6. Assuming that the roof drains are completely blocked during the PMP event, ove rflow scuppers limit the depth of water to within the design load carrying capability of th e roofs. Those safety-rel ated structures that do not have this relief capability structurally can carry the entire PMP accumulations.

2.4.3 PROBABLE

MAXIMUM FLOO D ON STREAMS AND RIVERS

Analyses for probable maximum flood (PMF) and SPF on the Columbia River (Reference 2.4-2) are consistent with the requirements of Regulatory Guide 1.59, Revision 2. The SPF for the Mid-Columbia Reach of the highly developed and regulated Columbia River is defined as 570,000 ft 3/sec (Reference 2.4-4). The unregulated SPF for the same reach is 740,000 ft 3/sec. The unregulated PMF at the si te is estimated to be 1,600,000 ft 3/sec (References 2.4-2 , 2.4-4 , and 2.4-5).

Adjustment of the flood profiles for the Hanford region reported in Reference 2.4-4 , results in a regulated PMF of 1,440,000 ft 3/sec and a water level of 390 ft at the Seismic Category II makeup water structure. This structure is not designed to function thro ughout the PMF but is designed for the SPF (unregulated) of 740,000 ft 3/sec.

Although assumed to exist for the purpose of flood hydrograph calcula tions, Ben Franklin Dam is not a federally authorized project. As originally planned it would have been a low head dam with only a negligible ef fect on extreme flood flows (Reference 2.4-6). The design basis flood for the CG S site area results from the PMP event on the adjacent drainage basin and not from fl ooding of the Columbia River.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-020 2.4-7 2.4.3.1 Probable Maximum Precipitation

The PMP event which was presented in the CGS PSAR was subsequently reevaluated in the preparation of the PSAR for WPPSS Nuclear Pr oject No. 1 (Docket 50-460). The analysis presented here is consistent with the latter document.

Precipitation in the vicinity of the site has been classified by the U.S. Weather Bureau, Reference 2.4-3 , as convergence precipita tion, orographic precipitation, and thunderstorm precipitation. The methodology for predicting the total amount of precipitation from each of these events, as given in Reference 2.4-3 , requires the adding toge ther of the convergence PMP and the orographic PMP to obtain a single pr ecipitation for a general storm. A separate analysis is then required for thunderstorms. Thunderstorms in the vicinity of the site can be locally very intense for short periods of time and hence, have the potential for causing serious flooding. The PMP for both a general storm and a thunderstorm were analyzed as given in

Chapters 6 and 5 , respectively, of Reference 2.4-3 for a 38.5 mile 2 basin at the site. This basin is shown in Figure 2.4-8 and is described in Section 2.4.3.3. The calculated general storm PMP results in a 24-hr a nd 48-hr precipitation of 7.9 in. and 10.1 in., respectively.

A thunderstorm PMP yields 9.2 in. in a 6-hr period. Therefore, the thunderstorm is considerably more severe. The thunderstorm PMP hydrograph is

Time Rain (hr) (in.) 1 0.6 2 1.6 3 5.2 4 0.9 5 0.5 6 0.4 Total 9.2

2.4.3.2 Precipitation Losses

Infiltration losses have been estimated in the vicinity of the sites as 1.5 in./hr (Reference 2.4-7). However, for the analysis below, an average antecedent moisture condition (Condition II as defined in Reference 2.4-8) was assumed. As e xplained in the following section, the 60-minute retenti on loss rate is 0.15 in./hr.

2.4.3.3 Runoff and Str eam Course Models

The drainage basin common to the reactor building and spray ponds is shown in Figure 2.4-8. The entire area drains to a broad channel that extends in a north-south direction for about 7 miles, and ranges from about 2000 ft to over a mile wide.

All plant structures are located on C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-8 high ground to the west of the channel. At a point about 2.8 miles south of the reactor site, the four-lane Department of Energy (DOE) highway crosses the drainage basin. The area above this section is 33.2 miles

2. To evaluate the effect of the PMP event on the plant area, the peak discharge at the highway crossing, 2.8 miles downstream of the plant, was calculated using the U.S. Bureau of Reclamation procedure for computing design floods on ungauged basins from thunderstorm rainfall in the western U.S. (Reference 2.4-8). Important assumptions used in the triangular hydrograph procedur e of Reference 2.4-8 are a. Hydrologic soil group B,

b. Land use and treatment cla ss - poor pasture or range,

c. Thunderstorm cover-index is brush-sage-grass combination with 50% or less cover density, and

d. Thunderstorm minimum 15-minute retention loss rate of 0.06 in./15 minutes

and 60-minute retention lo ss rate of 0.15 in./hr.

Additionally, no credit was taken in the hydrograph analysis for potential st orage in the stream channel or upstream sub-basins.

The time of concentration, T c , for the watershed above the hi ghway crossing was computed to be 7.5 hr. The PMF hydrograph is shown in Figure 2.4-7 for the 33.2 mile 2 drainage basin. A peak discharge of 21,400 ft 3/sec was determined.

Based on this PMF, an upstream water surface profile was determined using the Corps of Engineers HEC Standard-Step Procedure (Reference 2.4-9). A total of eleven cross sections were used (seven downstream, one at the plant, and three upstream as shown in Figure 2.4-3

). Details of the channel cro ss sections are shown in Figure 2.4-9. The Manning roughness coefficient was conservatively taken as n=0.035 in the main channel sections, and n=0.05 in the overbank areas.

Using the computational procedure of Reference 2.4-9, it was determined that the channel restrictions at cro ss sections 5 and 7 (Figure 2.4-3) do not control the flow. The stillwater elevation at the plant site (cross section 8) was determined to be 431.1 ft msl. The water surface profile is shown in Figure 2.4-10.

2.4.3.4 Probable Maximum Flood Flow

The PMF runoff hydrograph produced by the PMP at cross section 1 (Figure 2.4-3) is shown in Figure 2.4-7. The peak discharge at this location is 21,400 ft 3/sec.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-043 2.4-9 2.4.3.5 Water Leve l Determinations

As discussed in Section 2.4.3.3, the water elevation of a flood at the plant site generated by the PMP event is 431.1 ft msl. This flood condition has a higher estimated elevation than any flood of the Columbia River.

2.4.3.6 Coincident Wind Wave Activity Procedures published by the Corps of Engineers (References 2.4-10 and 2.4-11 were used to determine the wind wave activity. The effective fetch for the pr edominant July wind direction (north) is 3450 ft (0.65 miles). The effective fetch diagram is shown in Figure 2.4-11. The calculated extreme 2-year over water wind for th e north-to-south directi on, based on area data, is 63.5 mph. This wind results in a maximum wave he ight of 4.0 ft, with the assumption of a water depth of 12 ft (the average depth in cro ss sections 8, 9, and 10). The other potential wind directions ENE and ESE were eval uated but found to be less severe.

The wind setup has been computed to be 0.3 ft , and the maximum wave runup is 1.9 ft on a smooth, 1 on 8 slope of compacted naturally occurring sands and gravels. Therefore, the design water surface elevation is 433.3 ft msl. This is less th an the east spray pond overflow weir at elevation of 434.5 ft msl.

2.4.4 POTENTIAL

DAM FAILURES, SEISMICALLY INDUCED

Analyses of floods resulting fr om potential dam failures were investigated by the Corps of Engineers for the Columbia River. These studies are consistent with Regulatory Guide 1.59, Revision 2. The flood resulting from the breaching of Grand Coulee Dam is considered in lieu of a seismically induced flood.

In 1951, the Seattle District Corps of Engineers made a confidential study (now declassified) to determine artificial flood hydr ographs and the flood profile in the Columbia River Valley resulting from breaching the Grand Coulee Dam by enemy attack. The studies covered a spectrum of conditions in terms of breach openings and hydrologic conditions that might prevail at the time of attack. A postulated seismic failure of Grand Coulee Dam could result in displacement of part of the structure, but it would still act as a restriction or weir and minimize the hydraulic failure. For this re ason, the explosion-i nduced artificial flood represents an upper limit to se ismically induced failures. The failure of Grand Coulee Dam would initiate a catastrophic flood, which woul d be augmented by the failure of the earth portions of downstream dams and subsequent release of the storage pools.

Figure 2.4-5 shows water surface profiles for RM 323 to RM 358 for various river flows, including Artificial Flood No. 1. This flood provides a "limiting case" assessment of the conservatism of CGS

elevation. This flood would have an outfall peak of 8,800,000 ft 3/sec at Grand Coulee Dam at the moment of breaching and a peak discharge at RM 338 (Richl and) of 4,800,000 ft 3/sec.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-10 A base flow of 50,000 ft 3/sec was assumed above the mout h of the Snake for this flood (Reference 2.4-12). An arbitrarily assumed dramatic failure of Arrow and/or Mica Dams in Canada could result in greater releases of storage in terms of volume than that fr om the Grand Coulee Dam, but the effects of such postulated releases are mitigated by a combination of valley storage and critical (flow limiting) valley cross sections. The Corps of Engineers states (Reference 2.4-13) that the river channel restrictions at Trail, British Columbia, would restrict river flow to about 3.1 x 10 6 cfs, regardless of the postulated dam failure. A major failure upstream would result in this maximum flow for many days causing overtopping of Grand Coulee Dam. An analysis by the Bureau of Reclamation (Reference 2.4-14) concluded that overtopping which might result from the failure of upstr eam dams will not cause failure of either the Grand Coulee Dam or the Forebay Dam.

Various studies (References 2.4-12 , 2.4-15 , 2.4-16 and 2.4-17) made by the Corps of Engineers, and others, since 1951 have considered that breaching of Grand Coulee Dam would represent the worst catastrophic event for downstream Columbia River occupants.

Although these studies be ar no relationship to flooding from natural causes, they have been used as the basis for a very c onservative, limiti ng case approach.

Figure 2.4-5 shows water surface profiles for RM 323 to RM 395 for artificial and real stage flows, one of which corresponds to Artificial Flood No. 1 noted earlie r, which has been established (Reference 2.4-18) as conservative (limiting case) criteria for Co lumbia River flooding. Since the base flow used to develop these curves was 50,000 ft 3/sec, an additional 570,000 ft 3/sec is added to account for simultaneous occurrence of the regulated SPF.

2.4.4.1 Dam Failu re Permutations

The effect of potential dam failure on the water levels at the site is determined using the following assumptions:

a. The Columbia River is at flood stage, with a SPF (570,000 ft 3/sec regulated);

b. The reservoirs in each storage pool are full;

c. A massive hydraulic failure occurs at Grand Coulee Dam, releasing 8,800,000 ft 3/sec; d. Following the above assumed failure, a ll downstream dams between CGS site and Grand Coulee Dam suffer some degree of failure and release their storage reservoirs to the flood. [The result of a stability analysis (Reference 2.4-15) showed that all mass conc rete portions of the dams will resist sliding and overturning with the possible exception of part of Rock Island Dam.];

C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 2.4-11 e. The explosion-induced failure of Grand Coulee Dam represents a more severe failure than any seismic event b ecause of the failure mechanism;

f. Failure of Arrow and/or Mica Dam could result in gr eater release of storage volume than Grand Coulee Dam; however , the peak flow is limited to 3,100,000 ft 3/sec due to channel restrictions at Trail, British Columbia; and

g. Overtopping of Grand Coulee Dam would occur with failure of Arrow and/or Mica Dams in Canada. The failure of Grand Coulee, as a result of overtopping, is not considered to be a credible event in view of its concrete c onstruction and rock abutments.

2.4.4.2 Unsteady Flow Analysis of Potential Dam Failures

The flood hydrographs developed by the Corps of Engineers are based on the results of extensive studies of the ph ysical characteri stics of the flood route (References 2.4-12 and 2.4-15). Subsequent studies made by the Co rps of Engineers ver ify these results (Reference 2.4-17). Water levels following such a flood would depend on the status of reservoir storage upstr eam from Grand Coulee Dam but, w ithout regulation of some dams, would approximate the natural seasonal flow conditions.

2.4.4.3 Water Level at Plant Site

The water elevations associated with lim iting case flood (LCF) levels are shown in Figure 2.4-5. RM 350 provides the control for back water flow to the plant area which is sheltered by higher ground eas t of WNP-1 and WNP-4.

Elevation at RM 350 (dam breach flood = 422 ft msl 4,800,000 ft 3/sec plus SPF, 570,000 ft 3/sec)

Allowance for simultaneous wind

and wave action + 2 ft Elevation: 424 ft msl=LCF An adequate margin exists between the resulta nt flood elevation and th e plant elevation of 441 ft msl.

2.4.5 PROBABLE

MAXIMUM SU RGE AND SEICHE FLOODING

The location of the CGS site is not close to any water body whic h experiences seiche flooding.

Thus the site is not vulnerable to such flooding.

C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 2.4-12 2.4.6 PROBABLE MAXIMUM TSUNAMI FLOODING

The location of the CGS site is in south-central Washington and it is not adjacent to any coastal area. It is not, therefore, vulnerable to tsunami flooding.

2.4.7 ICE EFFECTS

Historically, the Columbia River has never experienced complete flow stoppage or significant flooding due to ice blockage. Periodic ice blocking has caused reduced flows and limited flooding for only relatively s hort periods of time.

The most significant icing in recent recorded history occurred during the winter of 1936-37 prior to the construction of the

upstream regulating dams. A relatively thick sheet of ice form ed across the river. The minimum flow recorded near the Priest Rapids Dam site during this winter was 20,000 cfs. However, the ice forming on the river was caused primarily by the low flow rather than the reverse. The deltaic mouths of many of the tributaries to the Columbia River are frequently blocked by ice causing backup of flood waters.

No instance of complete stoppage is known to have occurred.

Ice blockage is most likely to occur when water temperatures are already low, when flows are small, and when a significant cold spell occurs. With the completion of Grand Coulee and other dams on the Columbia River main stream , the seasonal temperature and flow cycles have drastically changed. These changes will further aid to reduce the intensity and timing of the conditions which may contribute to potential ic e blockage and flooding situations. Also average river flow rates, duri ng the winter months, have been increased significantly. The water temperatures have shown a shift in tim e such that the peak temperatures occur 30-45 days later than formerly. In addition, the low extreme temperatures measured have risen over the years.

The long term trends of temp eratures in the Columbia Ri ver been studied (Reference 2.4-19) using a 37 year record of measured temperatur es. The trends for the maximum, average and minimum temperatur es are shown in Figure 2.4-12. The erection of dams on the upper Columbia River has caused the extreme high a nd low river temperatur es measured at Rock Island Dam (Columbia RM 453, 101 miles above the CGS site) to converge toward the average. Winter water temperatures are c onsiderably warmer and summer temperatures cooler with a slightly lowe red average of less than 1

°F occurring during the 37 years.

On the basis of these studies and the recorded observation of 25 years of operation of the Hanford plutonium production plant s, it is concluded t hat the potential for ice blockage or the combination of blockage and flooding behind i ce dams is so low as to be considered insignificant. The erection of Mica, Arrow, and Libby Dams in the Columbia River Basin headwaters is expected to further raise winter water flows and also to increase winter water temperatures somewhat.

C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 2.4-13 In any event, ice flooding will not effect the capability to shut down the reactor in a safe and orderly manner. Also, the daily fluctuating st age of the river at the intake location will discourage formation of sheet i ce as well as ice jams. Ice fl ows, should they occur, will normally pass over intake structure due to rela tively high wi nter discharge in the river.

2.4.8 COOLING

WATER CANALS AND RESERVOIRS

There are no cooling water canals. The two spray ponds located southeast of the reactor building designed as Seis mic Category I structures, have rein forced concrete side walls, and reinforced concrete base mats at el. 420 ft msl. The finished grade at the spray ponds is approximately at el. 434 ft msl a nd have top of wall elev ations of 435 ft ms

l. The spray ponds are the ultimate heat sink for normal re actor cooldown and for emergency cooling.

The spray ponds are a part of the standby service water system which is discussed in Section 9.2.7. See also Section 2.4.11.6.

During normal reactor operation, the cooling water necessary for the plant is supplied from the cooling tower basins.

2.4.9 CHANNEL

DIVERSIONS

The Columbia River flow in the Hanford reach is controlled to a large extent by regulation of the upstream reservoir projects. The riverbed in the vicinity of th e site is well defined and it is very unlikely that the riverbed would be diverted from its present location by natural causes.

Any possible effect on water supply to the ma keup water pump house from riverbed changes would come from extremely slow changes which can be corrected if a nd when they occur.

As discussed in Section 2.4.7 , the river has not frozen over in Hanford reach during at least the past 25 years, and icing on th e river has not been a problem at pump house or outfall structures associated with th e plutonium production plants.

2.4.10 FLOODING PROTECTION REQUIREMENTS

The design considerations of sa fety-related facilities to withstand floods and flood waves are described in Section 2.4.2.2. The PMF is discussed in Section 2.4.3.

All safety-related facili ties are housed in Seismic Category I structures protected from flooding and designed to withstand the static and dy namic forces of all postulated floods. Flood considerations are de scribed in Section 3.4 and the design of Seismic Category I structures, for all conditions including flood, is described in Section 3.8.

In the event of a flood at the site, it will be possible to place the plant in a safe shutdown condition.

C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 2.4-14 All non-safety-related fa cilities with the excepti on of the makeup water pump house, are above the LCF elevation. The flooding of the makeup water pump house would not affect safety-related equipment and woul d not affect the safe shutdown of the plant. The approximate finished grade at all Seismic Category I structures except the spray ponds is at elevation 440 ft msl. The finished grade of the spray ponds is 434 ft msl.

The PMF elevation of the Columbia River (described in Section 2.4.3), at the site, is estimated to be 390 ft msl.

Seismic Category I structures ar e designed to withstand the st atic and dynamic forces which could result from a flood due to a breach of Grand Coulee Dam. Since this represents the LCF, the structures are also considered secure against the forces due to the lower PMF.

The access openings to all seismi c Category I structures are locat ed well above all flood water elevations, including that due to wind and wave action.

2.4.11 LOW WATER CONSIDERATIONS

As described in Section 2.4.1.1 , plant water needs are supplied th rough an intake structure in the Columbia River. The top of the makeup wa ter intake screens (at RM 352) are set below the water surface elevation that would be associated with the minimum allowable flow (36,000 cfs) at the federally licen sed Priest Rapids Da m (at RM 397). Water levels at the CGS intake are not influenced by backwater from the downstream McNary Dam (RM 292). The Columbia River Basin upstream of CGS has in excess of 35 million acre-ft of usable reservoir storage capacity. Because of this storage and highly regulated ri ver flows, it is improbable that flows below the licensed minimu m will occur. Based on data for 1961 throu gh 1994, 7-day low flow with a recurrence in terval of 100 years has been estimated at 44,500 cfs.

Even if some event (e.g., very severe dr ought) caused the makeup water system to be inoperable, the loss of water would not compromise the safe s hutdown of the plant. As is discussed in Sections 9.2.5 and 9.2.7 , shutdown cooling water is supplied by the ultimate heat sink which contains a 30-day supply of water in two spray ponds. The only scenario in which the makeup water pump house is called on to supply water in an emergency situation is when a tornado removes a significant quantity of spray pond water (see Section 9.2.5.3). Therefore, the low river water condition is not a situation re quiring safety-related fe atures and procedures.

2.4.11.1 Low Flow in Streams

Reservoir projects on the Columb ia River Basin upstream of th e proposed site have a total usable storage capacity in excess of 35 million acre-ft. This capacity is sufficient to maintain a flow in the Columbia River, at the proximity of CGS, of 36,000 ft 3/sec for over 1 year with absolutely no inflow from other s ources. Because of this regul ation, the anticipated minimum C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-15 and maximum monthly mean flow rate s will be 60,00 0 and 260,000 ft 3/sec in the vicinity of the site. It is improbable that minimum flows below that administratively set for dam operation (36,000 ft 3/sec) will occur due to drought conditions.

Columbia River st orage measurements have been extrapolated down to 25,000 cfs and are shown in Figure 2.4-13. The river elevation at RM 352, site of the CGS makeup water pump house, is 341.73 ft msl and has a corresponding flow of 36,000 ft 3/sec. 2.4.11.2 Low Water Resulting From Surges, Seiche s, or Tsunami There exists no possibility of lo w water conditions resulting from meteorological or geoseismic generated surges, seiches, or tsunami unless such natural phenom ena effected rapid closure of the Priest Rapids Dam, which is located 45 miles upstream fr om the proposed site. Rapid closure of the dam would cause a negative surge to be generated downstream.

A complete stoppage of flow is an unlikely e vent because of the redundant equipment and

operational procedure in place at the dam. Provisions to guard against an accidental shut off of Priest Rapids Dam include:

a. A gate actuation button in the control room of the Dam which is used to maintain at least minimum licensed flow from the facility in th e event of one or more turbine shutdowns.

b. Independent motors on each gate which have redundant wiring and power supplies.

c. Electrical heating on four of the gat es to prevent ice buildup which might interfere with gate operation.

d. Multiple offsite power sources in additi on to an on-site diesel generator power backup for gate operation.

In the event of a rapid and complete stoppage of flow over Priest Rapids Dam the effect of the negative surge would pass the site in a few hours. Since the Priest Rapids Dam is a run of the river dam with low storage capacity, it is unlikel y that its closure can restrict th e Columbia River flow for a significant period of time before being topped.

2.4.11.3 Historical Low Water

Historical records of the U.S. Geological Survey gauging station (RM 394.5) located 2.6 miles

downstream from Priest R apids Dam show low daily averaged flows of 20,000 ft 3/sec (January 31, 1937) and low monthly averaged flows of 20,900 ft 3/sec (February 1937). An instantaneous low flow of 4120 ft 3/sec occurred February 10, 1932, due to activit ies connected with dam regulation of the rive r near Wenatchee, Washington, be fore construction of Priest C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-16 Rapids Dam. After completion of the Dam in 1956, the minimum flow rate of the Columbia River at RM 352, approximate location of the CGS makeup water pump house site, is 36,000 ft 3/sec. The flow is maintained by the Grant County PUD as operator of the Priest Rapids Dam (RM 397) under FPC license which states:

"The licenses shall so regulate the flow from the Project 2114 that it will not result in flows of less than 36,000 cfs of water at the Hanford Works of the Atomic Energy Commission except when conditions are beyond the licensee's control."

In eighteen years of operation the flow has not dropped below the specified minimum.

The annual average flow of the Co lumbia River below Priest Rapi ds Dam is in the range of 115,000 cfs. The effect of us e of upstream water for irrigat ion development on the stream flow has been taken into account and the modified me an monthly discharge vari ations for the period 1928-58 are shown in Table 2.4-4 and Figure 2.4-14. The discharge for the base period of 1929-58 was adjusted to reflect 1970 levels of water utilization, including water consumption due to activities of fl ood control, power genera tion and irrigation.

Figure 2.4-15 shows the exceedance frequency for annual low flows for the period 1929 through 1958 with 1970 conditions measured at the gauging station (RM 394.5) immediately below Pr iest Rapids Dam.

Because of the flow regulation on the Columbia River, the anticipated minimum and maximum monthly mean flow rates will approximate 60,000 and 260,000 cfs in the vicinity of the CGS site. The variations of the rive r flow in this reach are due not only to seasonal fluctuations, but also to the daily regulation of the power produci ng Priest Rapids Dam. Flow rates during the late summer, fall and winter may vary from a low of 36,000 cfs to 160,000 cfs each day.

The dependable yield for flows in the Columbia River below Priest Rapi ds Dam for periods of one year through 10 years, as well as the 30-year period 1929-58 is illustrated in Table 2.4-5. The flow duration curve resulting from a plot of Table 2.4-4 is shown by Figure 2.4-14 (Reference 2.4-20). This figure illustrate s the percentage of time equaled or exceeded for different amounts of flows below Priest Rapi ds Dam on a monthly and on an annual basis.

2.4.11.4 Future Controls

Flows in the Columbia River at Hanford are required to be maintained above 36,000 ft 3/sec. This is the licensed minimum flow of Priest Rapids Dam and, as such, is a parameter closely monitored and controlled by the Grant Co unty PUD. The Stat e of Washington has administratively set higher average daily minimum flows (greater than 40,000 ft 3/sec) and will attempt to have the FERC li censes for the dams modified to insure the minimums (Reference 2.4-21).

C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 LDCN-04-052 2.4-17 2.4.11.5 Plant Requirements

All cooling water is supplied to the plant cooling towers via the circulating water system, the plant service water system, or the standby se rvice water system described in Sections 9.2.1.2 and 10.4.5. In the event of an incident rendering th e cooling towers inoper ative, cooling water is supplied from the spray ponds by the sta ndby service water system, described in Section

9.2.7. These

are closed loop systems and the only water loss is through evaporative cooling.

Makeup to the plant cooling towers and spray ponds comes from the Columbia River. Should this capability be lost, the cooling load is taken over by th e spray ponds. These ponds have sufficient capacity to provide shutdown cooling water for 30 days without makeup. Other sources of water are availa ble to provide makeup after the initial 30-day period (see Section 9.2.5). Therefore, variation in river flow will not have any a dverse affect on the capability to shut down the reactor in a safe and orderly manner.

2.4.11.6 Heat Sink Dependability Requirements

At the minimum river flow of 36,000 ft 3/sec described in Section 2.4.11.3 , there is still sufficient submergence at the makeup water pu mps to provide full makeup water requirements at full power operation. Sump level indication and low level alarms are provided in the main control room. Should the sump water elevation fall below the minimum submergence level for the makeup pumps, due either to low river flow or blocked inlets, the plant would be shut down if the situation could not be readily corre cted with the safety-related standby service water coming from the spray ponds.

Section 9.2.5 discusses the design bases used in desi gning the two spray ponds which serve as the ultimate heat sink for CGS. Design of the CGS ultimate h eat sink is in compliance with the guidelines presented in Regulat ory Guide 1.27 Rev. 1, "Ultimat e Heat Sink for Nuclear Power Plants," dated March 1974. Th e CGS spray ponds serve as th e suction source and discharge point for the standby service water system.

This system is discussed in Section 9.2.7 and identifies the uses and quantities of wate r drawn from the ultimate heat sink.

2.4.12 DISPERSION, DILUTION, A ND TRAVEL TIMES OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS IN SURFACE WATERS Small amounts of liquid radioactiv e wastes, processed within the plant and containing traces of radioactive nuclides, are discharged ultimately to the Columbia River via the plant blowdown line as discussed in Section 11.2.2.2.6 (see Figure 2.4-16

). In the vicinity of CGS, the Columbia River is wide, relatively shallow, and fast flowing.

Field measurements have shown river velocities near the CGS discharge to be about 3 ft/sec for minimum flows (Reference 2.4-22) and 4.5 ft/sec for aver age flows (Reference 2.4-23). At the point of discharge the river is about 5 ft deep at minimum flow. Based on a dye dispersion study (Reference 2.4-22), the local eddy diffusivity at low flow has been conservatively estimated to C OLUMBIA G ENERATING S TATION Amendment58 F INAL S AFETY A NALYSIS R EPORT December2005 2.4-18 be 4 ft 2/sec (Reference 2.4-24). With a combination of minimum river flow and maximum blowdown, it is estimated that an effluent woul d be diluted by a fact or of about 60 at a distance of 300 ft and a fa ctor of 200 at 3000 ft. Dilution factors and travel times for calculating doses to downstream water users are di scussed in the CGS Offsite Dose Calculation Manual (ODCM).

Downstream surface water us ers are listed in Section 2.4.1.2. The travel time to the nearest withdrawal which could be affect ed by an accidental release is approximately 1 hr. At that point a radioactive release would be essentially completely mixed with the river resulting in a dilution factor of 1:200,000. It is concluded that water users ar e sufficiently removed from the release point, and the Columbia Ri ver is sufficiently di spersive to preclude adverse impacts due to accidental releases. The di spersion characteristics of the river and the effects of routine releases are discus sed in Sections 5.1 and 5.

2 of the Environmental Re port - Operating License Stage.

2.4.13 GROUNDWATER

2.4.13.1 Descripti on and Onsite Use

Subsurface soil conditions, across the site , have been class ified as follows:

a. Loose to medium dense, fine to coarse sand with scattered gr avel (glaciofluvial sediments).

b. Very dense, sandy gravel with interbedded sandy and silty layers (Ringold Formation, Middle Member).

c. Very dense, interbedded layers of sandy gravel silt and soft sandstone (Ringold Formation, Lower Member).

d. Basalt bedrock which forms the bedrock beneath the area.

The lithologic character and water bearing properties of the geologic units occurring in the Hanford region are summarized in Table 2.4-6. In general, groundwater in the surficial sediments occurs unconfined, alt hough locally confined zones exist. Water in the basalt bedrock occurs mainly under confined conditions. Occasionally, the lower zone of the Ringold Formation occurs as a confined aquifer, separated from the overlying unconfined aquifer by thick clay beds which possess a distinct hydraulic potential.

The unconfined aquifer consists of both glaciofluvial sand and gravel deposits and the Ringold silts, clays, and gravels. Since these materials are very heterogeneous, often greater lithologic differences occur within a given bed than between beds. In the vi cinity of CGS the water table is below the top of the Ringold Formation (see Figures 2.5-64 and 2.5-65). The unconfined C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-19 aquifer bottom is the basalt bedrock in some areas and silt/clay zones of the Ringold Formation in other areas. Clearly the bottom of the unconfined aquifer is not a continuous lithologic surface.

The Hanford Reservation contains over 2200 wells cons tructed from pre-Hanford work days to the present). Approximately 600 of these wells are used for gr oundwater monitoring (Reference 2.4-25). Figure 2.4-17 identifies the well locations in the Hanford Reservation as of September 1975.

Figure 2.4-18 shows the December 1975 groundw ater contour map. In general, the groundwater gradient resulting from groundwater flowing under the Reservation is the highest in the southwestern area toward Rattlesnake Mountain, and slopes toward the Hanford 200 Areas near the center of the reservation. From the 200 Areas the general slope in the gradient is toward northeast and southeast.

A groundwater contour map base d on the potential construction of the Ben Franklin Dam at approximately RM 348 is illustrated by Figure 2.4-19. The CGS design basis groundwater level is based on the possible construction of the Ben Franklin Dam and is taken to be 420 ft msl, whereas the most recent study indi cates that the water table would be about 405 ft msl (Reference 2.4-26). The feasibility of constructin g Ben Franklin Hydroelectric Dam has been extensively studied.

Its proposal was strongly c ontested by local groups and individuals concerned with environmental protection and preservation. Additionally, the matter of the impact such a facility would have on the DOE Hanford Reservation was believed by some to preclude its construction. Finally, the cost

/benefit ratio was belie ved by many to be too low to make the project viabl

e. The combination of the unresolved impediments to the project has effectively, though not conclusively, relegated it to a very low priority status.

Planning studies for the project by the Corp s of Engineers were suspended in 1969 and reinitiated in October 1979 as part of the development of a m anagement plan for the Hanford reach. The most recent studies were terminated in November 1981.

Impermeable groundwater boundaries are the Rattlesnake Hills, Yakima Ridge, and Umtanum Ridge on the west and southwest sides of the Ha nford Reservation. Gable Mountain and Gable Butte also impede the groundwater flow, as well as other sma ll areas of basalt outcrop above the water table. The Yakima River recharges the unconfined aquifer alon g its reach from horn Rapids to Richland. The Columbia River fo rms a hydraulic potential boundary which is a discharge boundary for the aquifer.

The major source of natural recharge is precipitation on Rattlesnake Hills, Yakima Ridge, and Umtanum Ridge.

Minor changes would be expecte d in the groundwater elevati ons during the summer months because of the charging stage of the Columbia River, which hi storically reac hes peak flood stage in June. Because CGS is located about three miles from the river and because of the permeability characteristics and enormous volume of the Ringold Formation, there is a substantial time lag in changing water levels.

For the same reasons, th e range in water table fluctuations is very small.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-20 Natural recharge due to precip itation over the lowlands of th e Hanford Reservation is not measurable as the evaporation potential during the summer mont hs greatly exceeds total precipitation. Data on migration of moisture from natural preci pitation in deep soils (below 30 ft) show movement rates less than 1/2-in./y r at one measurement site (References 2.4-27 , 2.4-28 and 2.4-29). The major artificial recharge of ground water to the unconfined aquifer occurs near the Hanford 200 East and 200 West Areas. The la rge volume of process water (1.35 x 1011 gal) discharged to ground during 1944-1973 has caused the formation of significant groundwater mounds in the water table (Figures 2.4-20 and 2.4-26). Other local groundwater mounds formerly existed along the Columbia Ri ver. The present Hanford 100-N Area mound is the only one of these remaining. A minor recharge mound also exists at the Hanford 300 Area.

The unconfined aquifer is charac terized by its hydraulic conductiv ity, the storage coefficient, and the effective porosity. The hydraulic conductivity relates th e water flow quantity to the hydraulic potential gradient, while the effective porosity gives the fr action of por ous media volume that is available to transmit ground water flow. The storage coefficient relates a change in the water table elevation to a change in the volume of water contained in the aquifer per unit horizontal area. In the limit of no delayed yield, the storage coefficient is equal to the effective porosity of the soil through which the water table moves. These parameters vary widely over the H anford Reservation.

Qualitatively the hydraulic conducti vity, storage coefficient, and e ffective porosity distributions are a function of the different geologic formati ons in the unconfined aquifer. Ancestral Columbia River channels which incised in the Ringold Formation are now filled with more permeable glaciofluvial sedime nts. These channels have been identified extending eastward along the northern and southern flanks of Gable Mountain and extending southeastward from the 200 East Area to th e Columbia River (see Figure 2.4-21

). These permeable channels are reflected in the groundwater flow pattern of the region.

Quantitative measurements of the hydraulic conductivity of the unconfined aquifer have been made on the Hanford Reservation using a variety of technique s: pumping tests, specific capacity tests, and tracer tests. The most common method has been the pumping tests. Values obtained for the Ringold Formation range between 10 to 650 ft/day with a median of about 130 ft/day. In sharp contrast ar e the very large hydraulic c onductivities of glaciofluvial sediments, ranging from 1, 200 to 12,000 ft/day (Reference 2.4-30).

The storage coefficient is much more difficult to measure in the field and estimates are, therefore, less common. For th e unconfined aquifer, estimates of the storage coefficient have ranged from 0.01 to 0.1 (Reference 2.4-30). An areal estimate of 0.11 has been provided for the 200 West Area based on the growth of groundwater mounds (References 2.4-30 and 2.4-31). The median specific yield (effective por osity) has been estimated by various researchers at Hanford to range from 4.8% to 11%; most commonly it is assumed to be 10%

(Reference 2.4-32).

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-21 The unconfined groundwater aquife r is characterized by the contour map of the hydraulic potential or water table. The map for December 1975 appears in Figure 2.4-18. The depth to the water table varies gr eatly from place to place, depending chiefly on local topography which ranges from less than 1 to more than 300 ft below the land surf ace. Beneath most of the Hanford 200 Area disposal sites the depth of the water table average s about 250 ft. The current estimate of the maximum saturated thickness of the unconfined aquifer is approximately 230 ft.

The chemical quality of the groundwater in the unconfined aquifer is measured at seven locations. Sodium, calcium, and sulfate ions are measured as well as pH. Chromium and fluoride ions associated with fuel manufac turing operations are anal yzed from Hanford 300 Area wells. Nitrate ion, which is a waste product from the manufacturing and chemical separation operations, is monitored over the entire Hanford Reserv ation. Annual maps of the nitrate ion concentration near the surf ace of the unconfined aquifer are published (Reference 2.4-33). The map showing nitrate concentration for December 1975 appears in Figure 2.4-22.

The radiological status of th e groundwater near the surfac e of the unconfined aquifer is monitored regularly (Reference 2.4-34) and reported annually. Plots of gross beta (ruthenium) plumes and the tritium plumes are shown in Figures 2.4-23 and 2.4-24 for December 1975 (Reference 2.4-33). Since the nitrate ion is not adsorbed in the soil it can be used as a tracer for groundwater movement. The extent of move ment of waste water containing radionuclides can thus be plotted. Respecti ve tritium and nitrate ion concen trations under the CGS site are currently ranging from 30 to 300 pci/ml and 4.5 to 45 mg/l depending on the sampling location. Concentration guide for drinking water is 3,000 pci/ml for tritium and the recommended drinking water standard is 10 mg/l for nitrate ions.

Gross beta concentrations do not extend to the site.

From the research that has been done to date, it appears that there are a number of confined aquifers underlying the Hanford Re servation. Relatively im permeable confining beds commonly include the individual basalt flows and the silts and clays of the lower part of the Ringold Formation.

Within the basalt sequence, groundwater is transmitted primarily in the interflow zones, either in sedimentary beds or in th e scoria and breccia zones forming the tops and bottoms of the flows (References 2.4-35 and 2.4-36). Basalt flows in the Pa sco Basin have been eroded particularly in the anticlinal ri dges. In some locations th e basalts are highly jointed and contain breccia, pillow and pl agonite complexes through wh ich groundwater can move.

Consequently, hydraulic potential differences between water bear ing zones in the upper part of the basalt sequence are small o ver hundreds of feet of depth. The lowermost Ringold Formation silts and clays are of variable thickness. Distinct hydraulic potential differences have been observed between aquifers below th e silts and clays and the unconfined aquifer.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-22 Groundwater flow in the uppermost confined aquife r is also to the southeast with possible discharge into the Columbia River somewhere below Lake Wallula. However, the flow rates are regarded as quite small due to the low transmissivity range of this water bearing zone. Groundwater in the lower confined aquifers does not appear to cross the major anticlinal divides that define the Pasco Basin.

The piezometric or hydraulic potential map for the confined zones above the basalt (Figure 2.4-25) was based on measurements made in 1970.

In general, the hydraulic potential observed in the confined aquifer zones above th e basalt is greater than in the overlying unconfined aquifer. The main exception is in the vicinity of the Hanford 200 Area recharge mounds which have raised the poten tial in the unconfined aquifer.

One recharge area that has been identified is from the Yakima River at Horn Rapids. The piezometric map in Figure 2.4-25 also suggests rec harge from the upper Cold Creek Valley with flow toward a potential trough under the Columbia River. The Columbia Basin Irrigation Project to the northeast and east, and the Columbia River behi nd Priest Rapids and Wanapum Dams to the northwest are other probable rec harge sites in both these areas the basalt is exposed and is covered by pere nnially saturated unconsolidated deposits. A site of possible

minor recharge exists adjacent to Gable Butte and Gable Mountain anticline near the center of the Reservation.

Only 90 wells on the Hanford Reservation have been drilled to basalt. Thus data on the confined aquifers in the basalt flows are limited and more would have to be gathered to fully characterize the confined aquifers.

The plant is located on glacioflu vial outwash sands and gravels which are about 50 ft thick.

Below this layer occurs very de nse gravel. Sandy gravel occu rs in a sequence approximately 200 ft thick which is assumed to be the middle member of the Ringold Formation. The lower member of the Ringold Formation consists of a very compact, in terbedded gravel, sand, silt, and clay and extends down to a depth of about 500-525 ft.

Basaltic bedrock underlies the lower Ringold member, at approximately 550 ft depth.

The water table is about 60 ft below the ground surface level at CGS. The water table elevation is about 378

+4 ft msl and appears to be stable.

The effective bottom of the unconfined aquifer is assumed to be at about 220-260 ft msl at the top of the lower Ringold Formation. Groundwater potentials from the lower Ringold and from the basalt water bearing zones are about 25 ft higher than that of the unconfined aquifer. Test borings down to 925 ft reveal there are water bearing zones in the lower basalt flows and sedimentary interbeds at CGS. Piezometric level in basalt is 10 ft above unconfined wa ter table and hence artesian.

Under the CGS site the unconfined groundwater is moving easterly toward the Co lumbia River, the nearest discharge boun dary. Studies of the uppermost con fined aquifer indicate that the C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-23 potential gradients at the proposed site are oriented in the same general direct ion as those of the unconfined aquifer.

Three water supply wells are located on the site.

Two shallow wells were constructed in the unconfined aquifer (at approximatel y 240 ft deep) and a third well penetrates a confined aquifer in the underlying basalt flows (at approximately 695 ft deep). Norm al water supply is from the river, and the deep well is maintained in the standby mode to provide supplemental makeup water for the potable and demineralized water sy stem as needed. Pump ing capability is about 250 gpm. The two shallow wells were used during construction.

2.4.13.2 Sources

Regional use of the unconfined a quifer occurs at two nearby lo cations. The first is at the DOE's 400 Area located about 3 miles sout hwest of the CGS s ite as shown in Figure 2.1-3. Groundwater to this construction site is supplied from two wells and is used for sanitary and operation purposes. Maximum expected usage rate is between 2 million and 2.5 million gal/month. No data is available on drawdown tests performe d on the FFTF water supply wells 699-SO-7 and SO-8. The second location of ground water use is the WNP-1/4 site about 1 mile east of CGS. Wate r is drawn from two wells fo r sanitary and potable water requirements.

Usage rate is approx imately 250,000 gal/month.

The two onsite wells which drew from the unc onfined aquifer (699-13-1A and 1B) are 234 and 244 ft deep. Drawdown tests for each well showed 22 and 91 ft of drawdown respectively, at pumping rates of 250 gpm and test durations of about 25 hr. These well s are no longer used. The third well (695 ft deep) is sealed from the unconfined aquifer and draws from confined

water in the basalt. Drawdown on this well was 163 ft at a pumping rate of 275 gpm with a test duration of 25 hr.

Water table contours in the vici nity of CGS can be seen in Figure 2.4-26. The aquifer is assumed to be isotropic, ther efore, flow occurs along instantane ous streamlines perpendicular to the equipotential contours. The groundwater flow is toward the discharge boundary at the Columbia River to the east of the site. The hydraulic potential gradient in this area is about 8-10 ft/mile in the unconfined aquife

r. As described in Section 2.4.13.1 , recharge and discharge of riverbank storage occur along the Columbia River with daily fluctuations superimposed on the seasonal varia tions in river stage. Hydrographs of wells in the vicinity of the plant site (Figure 2.4-27) show that riverbank storage is not detectable even in years of extreme spring runoff at the two wells that are about one m ile from the riverbank. Thus no seasonal reversability of the gradients driving the groundwater flo w occurs. In other areas of the Hanford Reservation, the seasonal fluctuati ons of groundwater le vels from riverbank recharge can be detected 3-4 miles inland from the riverbank.

During early studies of groundwater in the area (References 2.4-37 and 2.4-31) little information was obtained on specific features at the plant site. The water table for 1944 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-24 (pre-Hanford Work conditions) was interpolat ed using 1948-1952 observation well data (Figure 2.4-20) and showed a water table elevation of about 370 ft msl under the site. The potential gradient was interpolated (References 2.4-31 and 2.4-38) to be about 5-6 ft per mile toward the Co lumbia River.

The earliest wells in the vicinity (699-2-3 and 17-5) were drilled in 1950.

Their hydrographs, presented in Figure 2.4-27 , show the gradual rise of the water table to approximately 15 ft above pre-Hanford Operations elevations. The peak rise in 1972 for well 699-2-3 shown on Figure 2.4-27 is believed to be a measurement error. Other wells were drilled in 1958, 1961, 1962 and 1966. Their hydrographs appear in Figure 2.4-27. Wells 669-14-E6-T and E5T also show the gradual rise of the water table at their respective loca tions. Smaller apparent water table changes at the site between 1944- 1974 (see 2.4.13.1 and Figure 2.4-27 , well 699-14-E6-T and 699-20-E5-T) indicated a zone of relativel y lower hydraulic conductivity in this area.

Well 699-9-E2 is a deep well perforated in both the unconfined and lower confined aquifer zones. Its hist orical hydrograph (Figure 2.4-27) reflects a composite of confined and unconfined potentials with disc ontinuities caused by sanding in an subsequent maintenance operations. The 1962 peak of the hydrograph of well 699-20-E12 (Figure 2.4-27) is due to the influence of the high confined aqu ifer potential before the installa tion of piezometer tubes in this deep well. The 1972 peak, could be bank rechar ge from the high river stage of that year but the lack of previous res ponse to river stage makes th e measurement suspicious.

Well 699-10-E12, also located within one mile of the river, does not show seasonal bank recharge (Figure 2.4-27

). Over the past two years, a decrease in the rate of rise is evident.

Soil test borings and water supply wells drilled in conjunction w ith the CGS construction site, confirmed the present contouring in terpretation of the water table.

Recent data from boring at WNP-1 and WNP-4 are not reflected in the water table maps shown in Figures 2.4-18 , 2.4-19 , 2.4-26 , and 2.4-25.

The historical well hydrographs for the uppermost confined aquifer in the vicinity of the plant site are given in Figure 2.4-27. Well number 699-20-E12-P shows a rather rapid rise of the confined aquifer potential in 1962-

65. It has been postulated that this rise reflects recharge to the confined zones from irriga tion across the river in the Columbia Basin Irrigation Project. The hydraulic potential in the uppe rmost confined aquifer near th e plant site is presently about 390 ft msl, which is about 25 ft higher than the pot ential of the overlyin g unconfined aquifer.

The effects of the groundwater w ithdrawal at the site have been estimated to be local. No drawdown has been detected in the ne arest observation wells , numbers 699-17-5 and 699-9-E2. The latter well is perforated over multiple aquifers so it does not give a representative measurement of the water table elevation. The radi us of influence (d efined to be the radius at which a 0.1 ft draw down exists) of the CGS wells has been estimated to be about 3500-4500 ft. This is based on the ten months of high rate of withdrawal during compaction C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 2.4-25 operations taking into account the ambient water table gradient. The su bsequent reduction in withdrawal flow rate to 25% of the early value would shrink the radius of influence considerably.

There is no groundwater recharge area within the influence of the plant. The 60-ft depth from the land surface to the water table and the arid condition of sediments above the water table make it virtually impossible to detect any r echarge from precipitation over this area.

2.4.13.3 Accidental Effects

An evaluation of a possible radioactive liquid release is postulated due to the rupture of a 700-gal decontamination solution concentrator waste tank within the radwaste building (see Figure 11.2-1). The released effluent was then assumed to reach the soil environment outside the building and to percolate to the water table unimpeded. On entering the groundwater, the postulated radwaste release is dispersed, sorbed, decayed, and diluted along the potential groundwater pathway from the plan t towards the Columbia River.

In the unconfined (water table) aquifer, there are no down gr adient groundwater users between CGS and the Columbia River. However, the construction water needs at WNP-1/4 are supplied by the two deep wells that withdraw groundwat er from the uppermost confined aquifer downgradient from the CGS radwaste building. During operation of WNP-1/4, these wells will be maintained in a standby mode. The uppermost screens in these wells are about 240 ft below the ground surface in the lower Ringold Formation. The effective bottom of the unconfined aquifer is generally assumed to be at the top of the lower Ringold Formation or about 200 ft below the surface. Thus, in all likelihood, any liquid radioactive sp ill to the groundwater beneath the CGS radwaste building would travel through the unconfined aquife r towards the Columbia River. However, fo r conservatism, analyses of postulated radionuclide movement assume that the WNP-1/4 wells draw from th e unconfined aquifer. Th e remainder of this subsection provides estim ates of travel times of critic al radionuclides to move from the postulated spill to receptors and the corresponding concentration reduction factors.

For an assumed one-dimensional groundwater movement, the groundwater tra vel time, t, is the path length, L, divided by the groundwater velocity (seepage velocity), u. The groundwater velocity is the Da rcy (apparent) velocity divide d by the effective porosity, u = Ki/n e, where K is the lateral perme ability (hydraulic conductivit y) of the aquifer, i is the hydraulic gradient, and ne is the effective porosity of the aquifer material.

For computational purposes, a conservative value fo r lateral permeability of 500 ft/day was selected to represent the unc onfined aquifer located in the Ringold Formation beneath CGS (see Figure 2.4-21

). From 2.4.13.1 , effective porosity is taken 0.10. From Figure 2.4-26 , the gradient in the water table aquifer between the plant and the Columbia River is about 8 or 9 ft/mile, and is taken cons ervatively as 10 ft/mile.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-26 Using the above parameter values , groundwater velocities were computed to be 10 ft/day.

With path lengths of 3.4 miles to the river and 1.0 mile to the WNP-1/

4 wells, the respective travel times are estimated to be 5.2 years and 1.5 years.

Generally, the critical radion uclides of concern for a postu lated liquid radwaste spill are 3 H, 90 Sr, and 137 Cs. These three radionuclides are fairly representative in terms of sorption characteristics, of t hose found in liquid radwaste tanks, since tritium does not sorb onto soil particles at all, strontium is an intermediate sorber, and cesium str ongly sorbs to soil particles.

The half-life of tritium is 12.3 years, whereas those of 90 S r and 137 C s are 29.0 and 30.1 years, respectively.

The travel time, t i , for a particular radionuclide movi ng through groundwat er depends upon the velocity, u i , of the radionuclide t i = l/u i where the radionuclide velocity is

u i = r f u in which r f is the velocity reduction fa ctor attributable to sorption r = 1/1+b n k fd In this equation, b is the bulk density of the aquifer material, n is the total porosity, and K d is the equilibrium distribution coefficient for a particular radionuclide.

The bulk density and total porosity are furthe r related physically as b = R s (1 - n) where R s is the real specific gravit y or particle density of the solid particles in the aquifer media. The particle density, R s , for Hanford soils is usually taken to be constant at 2.65 gm/cm 3 , (Reference 2.4-39). The bulk density b, of Hanford soils has been determined to range from about 1.5 gm/cm 3 to about 1.75 gm/cm 3 , with a median value of about 1.65 gm/cm 3 , (Reference 2.4-40). For the median value of bulk densit y, the corresponding to tal porosity is about 0.377.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-27 Using the above value for bulk density and total porosity, radionuclide tr a vel time, ti, through the ground w ater beneath WNP

-1/4 can be expressed as

t i = (1 + 4.4 K d) t The following summarizes radionucli d e travel times (in years) to the WNP-1/4 wells (1.0 miles)

and to the C o lumbia River (3.4 miles):

Nuclide Half-life, years K d t i @ 1.0 miles t i @ 3.4 miles 3 H 12.3 0 1.5 5.2 90 Sr 29.0 10 67.0 230.0 137 Cs 30.1 100 660.0 2300.0 The radionuclide concentration at the point of wa ter use will be determ ined by the amount of decay, dispersion, and sorption on the aquifer media. The minimum concentration reduction

factor, CRF min , along the centerline of the contaminant plume from an instantaneous point source is given by (Reference 2.4-41) CRF = C C (KKK)2Vmin oxyz1/2()/4 32 t for an effluent volume, V, with a specific gravity of 1.0 and an initial concentration, C o , released to soil with dispersion coefficients, K x , K y , K z , in the x, y, and z directions, respectively. This expression neglects the phenomena of sorption and decay which will be considered later.

It is generally accept ed that the dispersion coefficien ts are proportional to groundwater velocity for unidirectional flow, i.e.,

Kuxyzxyz,,,, = where x,y,z , are constants called disp ersivities which are a functi on of the nonhomogeneity of the material. The range in dispersivities in homogeneous granular aquifers may approach 1000 cm (33 ft) (Reference 2.4-42). Substituting this relationship into the above expression for concentration reduction, and noting that travel time is determined by pa th length and velocity, results in CRF min = (4L) ()2V3/2xyz1/2 For the conservative condition of x = y = z = 1.0, then C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-28 C R F = (4 L)2V m i n 3/2 The concentration reduction factors at the W N P-1/4 wells and at the bank of the Columbia

River, due only to dispersion, are 9.1 x 1 0 4 and 5.7 x 10 5 , respec t ively, f o r the 700-gal concentrated waste tan

k. When sorption and decay a r e included, the concentration

reduction is given by (Reference 2.4-42) C R F = (4 L)3 / 2 ( )x y z 1/2 2V e t i in which is the radionuclide decay constant defined in terms of the half-life, T1/2, of a particular radionuclide as

= 1n 2 T 1/2 The concentration reduction f a ctor can be expressed as

C R F = C R F (e t )m i n i The exponential term accounts for t h e effects of sorption and decay.

The only effect of sorption on concentration reduction is to increase the t r avel ti m e, t h us allowing more time f o r decay. Concentration reduction factors (CRF) for the radionuclides listed were calculated for path lengths of 1.0 mile (to WNP-1/4 wells) and 3.4 mil e s (to Columbia River):

Nuclide CRF (1.0 mile)

CRF (3.4 mile) 3 H 1.0 x 10 5 7.7 x 10 5 90 Sr 4.5 x 10 5 1.8 x 10 8 137 Cs 3.7 x 10 11 5.8 x 19 28 The above factors, derived thro ugh the application of conserva tive parameters, are used in Section 15.7.3 to evaluate concentrations offsite. The consideration of the WNP-1/4 wells is especially conservative. Gr oundwater contaminati on from the 200 Areas which reached CGS over six years ago (see Section 2.4.13.1) has not been detected at WN P-1/4. This substantiates that the WNP-1/4 wells do not draw from the unconfined aqu ifer or, alternatively, the hydraulic conductivities are much less than assumed.

C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-13-050 2.4-29 It should also be noted that if Ben Franklin Da m were ever constructe d, the concentration reduction factors at the river bank would be even larger than those noted above. This would be true, because the groundwater gradient (thus, the groundwater velocity) would be decreased as shown in Figure 2.4-19.

2.4.13.4 Monitoring or Safeguard Requirements

Plant water systems result in releases to the ground at a number of locations. Sanitary wastewater is routed to a cen tral treatment system comprised of lined aeration lagoons and stabilization ponds. This treatm ent plant also receives wastes from the Plant Support Facility, WNP-1 and WNP-4, and the DOE's 400 Area. Periodically the trea ted effluent is discharged to percolation beds.

As discussed in Section 2.4.2.3 , the storm water drainage syst em discharges to lined ponds northeast of the plant (see add itional description in Section 9.3.3.2.3.1

). Such sources as water treatment filter backwash es, heating, ventilating, and ai r conditioning (HVAC) air wash units, and some building sumps an d floor drains (see Section 9.3.3.2.3) also contribute to flow in the storm water system. Periodic testing and flushing of the fire protection system and cleaning of the cooling towers and standby service water ponds resu lt in localized discharges of water to the ground.

Monitoring of groundwater and pl ant-related discharges to ground is perfor med as described in the ODCM.

2.4.13.5 Design Bases for Subsurface Hydrostatic Loadings

CGS does not employ permanent dewatering syst ems. Site groundwater conditions are presented in Section 2.5.4.6 and the design bases for subsurfa ce hydrostatic load ings are given in Section 3.4.

The design-basis groundwater elevation used fo r subsurface hydrostatic loadings is 420 ft msl and was predicated on the possibl e future construction of Ben Fr anklin Dam at RM 348. As noted in Section 2.4.13.1 , planning for the dam has been terminated. The same section notes that the water table beneath CGS would rise to less than 405 ft msl if the da m were to be completed. The actual wa ter table beneath the project is about 385 ft msl (see Sections 2.4.13.1 and 2.5.4.6). The design-basis groundwater le vel is adequate to account for seepage from the ultimate heat sink spray ponds or the rupture of any Seismic Category I or nonseismic pipe. As discussed in Section 3.8.4.1.5 , the two, 250-ft 2 reinforced-concrete spray ponds are designed to Seismic Category I requirements and are designed to mitigate any possible water leakage. The bo ttom of the spray ponds are at 417 ft msl and the ponds are designed for external hydrostatic loading to 420 ft msl. The maximum combined leakage from the two ponds during the initial filling sequences was 120 gpm. It may be inferred from previous studies (Reference 2.4-7) that continued leakage cannot affect the groundwater level C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-30 beneath the ponds or other safety-related structures, th e closest of which (500+ ft away) is the diesel generator building with a foundation at 434 ft msl. These early CGS hydrologic studies evaluated the effect of cooling pond leakage and cooling tower blowdown at the project site.

Equivalent leakage/discharge rate s used in these studies were ve ry much greater than leakage from the spray ponds. For example, the con tinuous discharge of 2700 gpm of cooling tower blowdown to the depression just east of CGS wa s estimated to raise the water table about 20 ft beneath the point of discharge (Reference 2.4-7). Based on these previous studies, it can be concluded that the minimal amount of spray pond leakage will have no influence on the design-basis groundwater el evation of 420 ft msl.

With respect to a pipe break, th e 144-in. circulating water system pipe on the discharge side of the condenser would produce the maximum release of water. The maximum quantity of water released from such a rupture is 199,180 ft

3. This discharge would have a negligible (less than 1 in.) and temporary effect on the groundwater level. Failure of the pipe at its closest proximity to a safety-related bu ilding may result in temporary satu ration of the backfill. This material has been recompacted to a minimum relative density of 75% and an average relative density of 85%. As discussed in Section 2.5.4.8, these densities are not susceptible to liquefaction for motions associat ed with the safe shutdown ear thquake, and as discussed in Section 2.5.4.10 , temporary saturation would not significantly reduce the bearing capacity of the densely compacted backfill.

A continuous but undetected l eak from any major pipe would not influence the groundwater leve l enough to affect plant struct ures. This may be deduced from the time factors and water table rises predicted from a num ber of scenarios in the above mentioned hydrologic studies (Reference 2.4-7).

2.4.14 TECHNICAL SPECIFICATIONS AND EMERGENCY OPERATION REQUIREMENTS

The worst hydrological condition, as discussed in this section, is a flood caused by a postulated PMP event. This flood does not create an adverse hydrological condition on safety-related equipment. Emergency fl ood protection procedures are therefore unnecessary.

2.4.15 REFERENCES

2.4-1 Woods, V. W., "A Summary of Columbia River Hydrographic Information Pertinent to Hanford Works - 1894 to 1954," HW-30347, February 24, 1954.

2.4-2 Memorandum Report Colu mbia River Basin Lower Co lumbia River Standard Project Flood and Probable Maximum Flood , U.S. Army Engineers, North Pacific Division, Portland, OR, September 1969.

2.4-3 "Probable Maximum Precipitation," Hydrometeorological Report No. 43 , U.S. Weather Bureau (now NOAA), Northwest States, 1966.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-31 2.4-4 Corps of Engineers, Nort h Pacific Division, letter to V. C. St. Clair, Atomic Energy Commission, Richland O ffice, November 2, 1970.

2.4-5 Corps of Engineers, Nort h Pacific Division, letter to M. J. Hroncich, Burns and Roe Inc., February 7, 1972.

2.4-6 Corps of Engineers, Nort h Pacific Division, letter to M. J. Hroncich, Burns and Roe Inc., February 14, 1972.

2.4-7 Final Report on Hydrology Studies of the WNP-2 Site , July 1971, and Addendum I to Final Report on Hydrology Studies of the WNP-2 Site, by Battelle Northwest, Richland, WA, to Burns and Roe Inc., July 1971.

2.4-8 Design of Small Dams , U.S. Bureau of Reclamation, 1977.

2.4-9 "Water Surface Profiles," Vol. 6 Hy drologic Engineering Methods for Water Resources Development , U.S. Army Corps of Engineers, Hydrologic Engineering Center, July 1975.

2.4-10 Shore Protection Manual , U.S. Army Corps of Engineers, Coastal Engineering Research Center, 1975.

2.4-11 "Wave Runup and Wind Setup on Reservoir Embankments," Engineering Technical Letter No. 1110-2-221, Department of the Army, Corps of

Engineers, November 29, 1976.

2.4-12 Artificial Flood Possib ilities on the Columbia River , Corps of Engineers, Washington District, Washingt on, DC, November 20, 1951.

2.4-13 Rockwood, D. M, letter to R. Ch itwood, "Preliminary Estimate of Upstream Dam Failure Effects," June 21, 1972.

2.4-14 Arthur, H. G., letter to J. J. Stein, "Behav ior of Grand Coulee Dam and Forebay Dam When Subjected to Overtopping," August 1, 1972.

2.4-15 Artificial Flood Considera tions for Columbia River Dams , Corps of Engineers, U.S. Army Engineer Distri ct, Seattle, WA, August 1963.

2.4-16 Eugene Isaacson, Ca lculations of Severe Fl oods in a Developed River , January 1965.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-32 2.4-17 Artificial Flood Consider ations for Columbia River Below Chief Joseph Dam to Richland , Washington, U.S. Army, Corps of Engineers North Pacific Division, Portland, OR, January 1968.

2.4-18 Safety Evaluation by the Division of Reactor Licen sing, U.S. Atomic Energy Commission in the matter of Portland General Electric Company, Pacific Power and Light Company, City of Eugene, OR, Trojan Nuclear Plant, Docket No. 50-344, pp. 11-12.

2.4-19 Jaske, R. T., and S ynoground, M. O., "Effect of Hanford Plant Operation on the Temperature of the Columbia River, 1964-Present," BNWL-1345, Battelle, Pacific Northwest Laboratories, Richland, WA, November 1970.

2.4-20 "Columbia North Pacific Region Wa ter Resources Appendix V," Volume 1, March 1969.

2.4-21 Columbia River Instream Resource Protection Program, Department of Ecology, Olympia, WA, June 1980, pp. 61-67.

2.4-22 Vertical Mixing Characteristics of the Columbia River at RM 351.75, WNP No. 2 , Battelle Pacific Laboratories, Richland, WA, March 16, 1972.

2.4-23 Preoperational Environmental Monitoring Studies Near WNP-1, 2, and 4 August 1978 Through March 1980 , WPPSS Columbia River Ecology Studies Vol. 7, Beack Consultants, Inc., Portland, OR, June 1980.

2.4-24 Kannberg, L. D., Mathematical Modeling of the WNP-1/2/4 Cooling Tower Blowdown Plumes, Battelle Pacific Northwest Laboratories, Richland, WA, March 1980.

2.4-25 McGhan, V. L., and Damschen, D. W., Hanford Wells, PNL-2894, Battelle Pacific Northwest Laboratorie s, Richland, WA, May 1979.

2.4-26 Harty, Harold, "The Effects of Ben Franklin Dam on the Hanford Site,"

PNL-2821, Battelle, Pacific Northw est Laboratories, Richland, WA, April 1979.

2.4-27 Bierschenk, W. H., "Aquifer Characteristics and Ground Water Movement at Hanford," HW-60601, June 9, 1959.

2.4-28 Bierschenk, W. H., "Hydraulic Characteristics of Hanford Aquifers," HW-48916, Marc h 3, 1957.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-33 2.4-29 Honstead, J. F., McConiga, M.

W., and Raymond, J. R., "Gable Mountain Ground Water Tests," HW-34532, January 21, 1955.

2.4-30 Gephart, R. E., Spane, F. A., Leonhart, L. S., Palom bo, D. A., Strait, S. R., "Pasco Basin Hydrology", Hydrologic Studies within the Columbia Plateau, Washington - An Integration of Current Knowledge, (RNO-BWI-ST-5), by Rockwell Hanford Operations, Richland, WA, October 1979.

2.4-31 Newcomb, R. C., Stra nd, J. R. and Frank, F. J

., "Geology and Ground Water Characteristics of the Hanford Rese rvation of the U.S. Atomic Energy Commission," Professional Paper # 717, USGS., Washington, 1972.

2.4-32 Cole, C. R., and Reisenauer, A. E., "Variable Thickness Transient Model Assumptions and Boundary Conditions

," Battelle Pacific Northwest Laboratories, Richland, WA, August 1974.

2.4-33 Environmental Monito ring Report on the Status of Groundwater Beneath the Hanford Site, January- December 1975, BNWL-2034, Battelle, Pacific Northwest Laboratories, Ri chland, WA, January 1977.

2.4-34 Bramson, O. P. E., and Corley, J. P., "Hanford Environmental Surveillance Routine Program," Master Schedule -C Y 1973, BNWL-B-234, Battelle, Pacific Northwest Laboratories, Richland, WA, December 1972.

2.4-35 Newcomb, R. C., "S ome Preliminary Notes on Ground Water in the Columbia Basalt," Northwest Sciences , Vol. 33, 1, 1959, pp. 1-18.

2.4-36 LaSala, A. M., Jr., Doty, G. C., and Pearson, F. S., "A Preliminary Evaluation of Regional Gr ound Water Flow in Sout h-Central Washington,"

USGS Open File Report, January 1973.

2.4-37 Parker, G. G., and Pi per, A. M., "Geologic and Hydrologic Features of the Richland Area, WA, Relevant to Disposal of Waste at the Hanford -Directed Operations of the Atomic Energy Commission," Interior Report 1, USGS Report to Atomic Energy Commission , 101 pages, 5 Illus., 1949.

2.4-38 Kipp, K. L., and Mudd, R. D., "S elected Water Table C ontour Maps for the Well Hydrographs and Hanford Re servation, 1944-1 973," BNWL-1797, Battelle, Pacific Northwest Laboratories, Richland, WA.

2.4-39 Serne, R. J., Routs on, R. C., and Cochran, D.

A., "Experimental Methods for Obtaining PERCOL Model I nput and Verification Data," BNWL-1721, Battelle Pacific Northwest Laboratories, Richland, WA, 1973, p. 24.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 2.4-34 2.4-40 Rouston, R. C., a nd Serne, R. J., "Experimental Support Studies for the PERCOL and Transport M odels," BNWL-1719, Battel le Pacific Northwest Laboratories, Richland, WA, 1972, pp. 37, B-1, and B-2.

2.4-41 Carslaw, H. S., and Jaeger, J. C., Conduction of Heat in Solids , Oxford University Press, London, England, 1959.

2.4-42 Codell, R. B., and Schreiher, D. L., "NRC Models for Evaluating the Transport of Radionuclides in Groundwater," Proceedings of the Symposium on

Management of Low Level Radioact ive Waste, Pergamon Press, 1979, pp. 1193-1212.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.4-35 Table 2.4-1 Major Columbia River Basin Dams Location Dams River River Miles from Site Usable Storage 10 3 ac-ft Upstream Mica Columbia (Can) 666 12,000 Duncan Duncan 1,400 Arrow Columbia (Can) 429 7,090 Libby Kootenai 642 5,000 Hungry Ho r s e South Fork Flathead 3,160 Kerr Clark Fork 1,219 Albeni Falls Pend Oreille 483 1,153 Grand Cou l ee Columbia 245 5,200 Chief Joseph Columbia 193 -- Wells Columbia 164 117 Chelan Chelan 152 677 Rocky Reach Columbia 122 120 Rock Island Columbia 101 -- Wanapum Columbia 64 38 9 a Priest Rapids Columbia 45 17 0 a Downstream McNary Columbia 60 John Day Columbia 136 The Dalles Columbia 160 Bonneville Columbia 206 a Storage not availabl e for flood regulation.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.4-36 Table 2.4-2 Columbia River Tempe r atures N ear Columbia Generating Station MONT H L Y AVERA G E W A TER T E MP E RATU R E, IN C, AT RICHLAND, W A Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Annual Avera g e 1965 6.1 5.4 6.3 9.1 11.0 14.2 17.3 19.8 18.5 16.4 12.6 8.4 12.1 1966 6.9 6.2 6.8 10.3 12.1 13.5 16.2 18.8 19.4 15.6 12.6 9.5 12.2 1967 7.4 7.0 6.6 8.8 12.0 13.9 17.0 20.2 19.4 16.1 12.0 7.8 12.4 1968 5.7 5.0 6.0 8.8 12.8 14.3 17.0 18.7 18.3 15.0 11.4 7.4 11.7 1969 2.7 1.9 4.3 8.0 11.4 15.3 17.9 19.3 18.6 15.2 11.7 7.0 11.1 1970 5.3 4.9 5.7 7.9 11.7 15.4 19.0 19.9 17.5 14.9 10.6 5.9 11.6 1971 4.2 3.4 3.8 7.0 11.1 12.9 16.4 19.5 17.8 15.0 10.7 6.2 10.7 1972 3.3 2.2 3.7 7.0 11.0 13.3 15.5 18.1 16.9 14.0 10.5 6.1 10.1 1973 3.2 3.0 4.7 7.8 12.9 15.6 18.3 19.6 18.3 15.0 9.9 7.6 11.3 1974 3.2 3.2 5.2 8.2 11.3 13.7 17.4 19.4 18.8 15.4 11.5 7.9 11.3 Average 1965-1974 4.7 4.2 5.3 8.3 11.7 14.2 17.2 19.3 18.4 15.3 11.4 7.4 11.4 Minimum Daily 0.2 0.7 2.4 5.1 8.6 11.2 14.2 17.3 14.6 11.1 7.7 2.4 -- Maximum Daily 8.3 8.3 8.6 12.8 15.0 17.7 20.4 21.5 21.1 18.5 15.9 11.3 -- Recor d s since June 1964.

MONT H L Y AVERA G E W A TER T E MP E RATU R E, IN C, AT PRIEST RAPIDS DAM, WA 1961 5.4 4.7 4.7 7.4 10.4 13.7 17.3 18.9 17.8 14.9 10.4 6.6 11.0 1962 4.1 3.6 3.6 6.5 10.0 13.7 16.1 17.4 17.1 14.8 11.9 8.9 10.6 1963 5.3 3.8 4.6 6.5 10.4 14.0 16.6 18.4 18.3 16.3 11.9 7.7 11.2 1964 5.5 4.6 4.7 7.2 9.7 12.8 15.3 17.1 16.3 14.6 10.8 6.3 10.4 1965 4.4 3.3 4.1 6.6 10.0 13.3 16.1 18.4 17.3 15.3 11.9 7.8 10.7 1966 4.8 4.1 4.5 7.8 10.6 12.4 15.3 17.5 17.5 14.6 11.6 8.4 10.8 1967 5.9 5.7 5.0 6.8 10.1 13.3 16.1 18.5 18.2 15.4 11.3 7.2 11.1 1968 4.6 3.3 4.6 7.1 11.1 13.4 16.1 17.5 17.2 14.2 10.9 6.8 10.6 1969 2.4 1.5 3.4 7.2 10.8 14.6 17.1 18.2 17.7 14.8 11.5 7.6 10.6 1970 4.3 4.1 4.8 6.8 10.9 14.8 18.0 19.2 17.5 15.2 10.6 6.2 11.0 1971 4.0 3.5 3.6 6.6 10.7 12.6 15.3 18.4 17.2 15.2 11.3 6.8 10.4 1972 3.6 1.9 4.0 7.2 10.6 12.9 15.2 17.3 16.8 15.4 11.3 7.3 10.3 1973 2.3 2.9 4.8 7.7 12.5 15.4 17.6 18.8 17.8 15.2 10.3 7.7 11.1 1974 4.0 3.0 4.9 7.7 10.8 13.6 17.2 18.7 18.4 15.5 11.8 8.6 11.2 Average 1965-1974 4.0 3.3 4.4 7.2 10.8 13.6 16.4 18.3 17.6 15.1 11.3 7.4 10.8 Minimum Daily 0.3 0.3 2.2 4.3 7.5 10.6 13.1 16.6 15.3 12.2 7.7 2.3 -- Maximum Daily 7.6 6.2 6.9 10.1 14.6 17.1 19.3 20.2 20.0 18.7 14.4 10.5 -- Records since August 1960. Recorded values adjusted by computer-simulation to compensate for measurement errors and missing data.

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 0-0 0 0 2.4-37 Table 2.4-3 Downstream Surface Water Use r s Location of Diversion Approximate Miles Quantity Type Name Township Range Section Downstream (cfs) Use a Energy N o rt hwest 11 28 2 -- 90 IN Peter Kewit and Sons 11 28 2 -- 1 I L. L. Bailey 11 28 24 4 2 I H. D. Loyd 11 28 24 4 0.99 D,I Central Premix Concrete Company 11 28 27 4 2 IN Battelle Memorial Institute 10 28 14 8 4.4 I University of Washington 10 28 23 9 1.75 I City of Richland 10 28 24 9 0.67 D City of Richland 10 28 25 12 31 D City of Richland 10 28 25 12 23.25 D City of Richland 10 28 25 12 31 D City of Richland 10 28 35 12 93 D E. C. Watts 9 28 1 13 0.31 D,I H. S. Petty 9 28 1 13 0.48 I N. H. and M. E. Ketchersid 9 28 1 13 1.66 I G. C. Walkley 9 28 1 13 2.32 I R. T. Justesen, et al. 9 28 12 15 2.54 I Central Premix Concrete Company 9 28 12 15 1.10 IN City of Richland 9 28 13 17 2.0 I Benton County 9 29 28 19 1.0 I City of Kennewick 9 30 31 23 55.7 D City of Pasco 9 30 31 23 35.0 D F. J. Henckel 8 30 14 27 0.015 I Allied Chemical 8 30 14 27 3.55 IN Chevron Chemical 8 30 23 28 3.77 IN Chevron Chemical 8 30 23 28 40 IN Phillips Pacific Chemical Company 8 30 24 28 82 IN Phillips Pacific Chemical Company 8 30 24 28 20 IN

Boise Cascade Corporation 7 31 10 34 24.5 IN L. D. Hoyte, et al. 7 31 14 35 179.8 I D. Howe 7 31 23 36 6.4 I Crawford and Sons 6 30 27 47 32.8 I Barbarosa Farms 6 30 27 47 20 I Crawford and Sons 6 30 27 47 7.6 I Rainier National Bank 6 30 27 47 9.4 I Anderson and Coffin 5 29 5 49 242 I Horse Heaven Farms 5 29 6 50 82 I Horse Heaven Farms 5 29 6 50 550 I Horse Heaven Farms 5 29 6 50 290 I Anderson and Coffin 5 29 6 50 242 I a D - Domestic or municipal uses I - Irrigation and other agricultural uses IN - Industrial Includes only those water rights for which a permit or certificate has been issued C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.4-38 Table 2.4-4 Mean Disc h a rges in C F S of Columbia River Below Priest Rapids Dam, Modified to 1970 Condi t i ons Water Year Oct. N o v. Dec. J a n. F e b. Mar. Apr. May J une July Au g. Sept. Annual 19 2 8 22 4 000 10 9 900 90 9 00 19 2 9 82 3 00 78 4 00 10 1 100 10 3 000 10 8 000 72 6 00 85 2 00 62 0 00 71 3 00 87 6 00 97 0 00 94 3 00 86 9 00 19 3 0 87 9 00 89 8 00 10 2 700 93 5 00 90 7 00 83 1 00 72 5 00 81 7 00 90 2 00 98 8 00 97 6 00 92 7 00 90 1 00 19 3 1 86 8 00 89 6 00 10 0 000 82 2 00 90 8 00 88 4 00 74 5 00 81 7 00 10 4 000 10 2 200 99 4 00 85 8 00 90 4 00 19 3 2 87 4 00 88 7 00 10 2 000 95 0 00 10 9 200 77 8 00 90 7 00 15 7 500 15 6 700 74 6 00 97 8 00 90 6 00 10 2 300 19 3 3 89 6 00 69 7 00 10 2 700 12 8 800 16 7 100 97 9 00 11 8 900 18 5 900 19 6 600 18 0 300 12 1 900 10 0 200 13 0 000 19 3 4 10 0 600 10 4 200 12 8 000 13 9 600 20 3 400 19 6 700 24 3 100 22 1 200 16 8 800 10 4 500 10 0 000 10 1 000 15 0 900 19 3 5 82 0 00 72 4 00 10 9 200 13 2 100 13 2 000 11 1 300 11 7 600 14 7 500 15 6 900 13 1 100 99 3 00 96 9 00 11 5 700 19 3 6 90 2 00 86 2 00 10 7 900 11 9 400 79 8 00 80 4 00 81 5 00 16 0 500 12 3 300 83 4 00 93 2 00 89 4 00 99 6 00 19 3 7 87 6 00 87 5 00 10 5 400 96 6 00 10 0 600 84 4 00 63 5 00 70 4 00 76 9 00 87 8 00 10 2 500 91 5 00 87 9 00 19 3 8 89 3 00 83 1 00 88 7 00 11 1 000 12 4 100 86 8 00 11 0 700 14 2 400 14 6 800 15 4 100 90 4 00 89 2 00 10 9 700 19 3 9 83 4 00 77 1 00 91 7 00 12 7 200 90 4 00 83 0 00 10 8 500 10 0 000 11 2 400 95 5 00 96 9 00 90 8 00 96 4 00 19 4 0 85 8 00 85 4 00 90 5 00 13 3 200 98 0 00 89 2 00 11 0 700 89 7 00 10 1 700 94 1 00 96 0 00 91 6 00 97 2 00 19 4 1 84 3 00 79 6 00 92 5 00 99 4 00 92 2 00 87 9 00 13 7 400 76 9 00 73 2 00 84 0 00 91 5 00 88 7 00 90 6 00 19 4 2 96 0 00 82 7 00 91 4 00 11 4 100 11 9 000 84 6 00 11 5 900 10 5 300 14 8 400 10 1 400 10 2 000 88 3 00 10 4 100 19 4 3 87 9 00 65 8 00 86 8 00 10 5 600 15 0 600 11 6 000 13 2 400 20 2 600 13 4 300 14 7 700 10 1 300 88 9 00 11 8 300 19 4 4 81 3 00 77 2 00 96 8 00 99 3 00 11 0 600 78 7 00 88 2 00 88 0 00 69 1 00 81 2 00 94 6 00 84 4 00 87 4 00 19 4 5 90 1 00 90 9 00 10 3 600 88 5 00 94 0 00 86 5 00 77 8 00 11 2 800 67 8 00 88 8 00 99 3 00 87 4 00 90 6 00 19 4 6 86 2 00 85 7 00 92 5 00 95 6 00 11 7 700 90 8 00 11 2 200 17 8 100 17 0 900 13 4 500 94 4 00 91 1 00 11 2 500 19 4 7 79 6 00 81 3 00 93 1 00 11 6 000 13 7 800 13 5 200 15 5 900 18 4 400 16 3 400 13 6 300 89 9 00 85 5 00 12 1 500 19 4 8 94 7 00 96 0 00 11 3 900 11 3 200 20 2 800 16 6 700 13 7 700 19 3 400 25 7 600 19 4 700 12 2 900 10 1 900 14 9 600 19 4 9 88 0 00 83 6 00 97 7 00 12 6 000 11 4 000 80 0 00 12 3 000 16 6 400 18 1 600 82 7 00 92 2 00 87 8 00 11 0 200 19 5 0 79 0 00 69 5 00 10 6 800 12 3 300 15 5 200 14 5 400 13 6 400 19 7 500 20 0 200 21 1 900 11 4 800 96 2 00 13 6 400 19 5 1 91 8 00 87 9 00 10 2 600 11 5 400 22 3 400 18 6 200 19 5 600 18 8 800 17 1 300 17 4 300 11 0 300 91 7 00 14 4 900 19 5 2 94 2 00 98 8 00 11 2 200 12 6 500 15 5 200 11 3 300 13 4 600 17 2 400 13 5 800 14 5 100 88 7 00 85 9 00 12 1 900 19 5 3 85 5 00 83 9 00 10 3 900 95 5 00 12 4 800 87 8 00 98 7 00 17 4 000 16 8 300 14 1 400 99 0 00 89 2 00 11 2 700 19 5 4 83 6 00 89 8 00 11 0 300 12 2 100 15 3 600 13 5 200 12 4 200 19 1 200 22 4 900 22 8 400 16 3 600 11 4 400 14 5 100 19 5 5 98 7 00 10 3 400 12 6 600 13 2 400 14 3 900 10 2 700 11 0 500 10 4 300 18 1 800 19 3 300 11 1 900 91 0 00 12 5 000 19 5 6 95 7 00 94 5 00 97 0 00 10 8 100 20 6 500 20 0 600 17 3 500 24 5 800 21 2 600 20 0 400 10 3 600 90 7 00 15 2 400 19 5 7 87 4 00 82 9 00 10 9 400 13 2 100 14 5 100 10 1 200 11 3 600 18 2 700 17 6 500 12 0 900 89 0 00 86 9 00 11 9 000 19 5 8 77 5 00 75 2 00 83 3 00 12 0 200 12 3 700 10 7 300 12 5 000 17 2 800 17 2 900 Mean 87800 84700 101700 113200 132100 108600 119000 147900 147200 132800 102400 91800 114100

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.4-39 Table 2.4-5 Dependable Yield, C o l u mbia River Below

Priest Rapids Dam, Washington

Consecutive Years of Lowest Mean Flow Inclusive Years Lowest Mean Flow (cfs) Percent of 1929-1958 Mean 1 1937 86,600 75.9 2 1930-31 89,900 78.8 3 1929-31 92,900 81.4 4 1929-32 95,800 84.0 5 1937-41 96,400 84.5 6 1937-42 97,300 85.3 7 1936-42 98,400 86.2 8 1937-44 99,000 86.8 9 1937-45 97,900 85.8 10 1936-45 98,600 86.4 11 1929-58 114,100 100.0

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 2.4-40 Table 2.4-6 Major Geologic Units in the Hanford Region and Their Water-Bearing Properties System Series Geologic Unit Material Water-Bearing Properties Fluviatile and glaciofluviatile sediments and the Touchet

formation

(0-200 ft thick)

Sands and gravels occurring chiefly as glacial

outwash. Unconsolidated, tending toward coarseness and angularity of grains, essentially free of fines. Where below the water table, such deposits have

very high permeability and are capable of storing vast amounts of water.

Highest permeability value determined was

12,000 ft/day.

Pleistocene Palouse soil Wind deposited silt. Occurs everywhere above the water table. Quaternary Ringold formation

(200-1200 ft thick) Well-bedded lacustrine silts and sands and local beds of clay and gravel.

Poorly sorted, locally

semi-consolidated or cemented. Generally

divided into the lower

"blue clay" portion which contains considerable sand

and gravel, the middle conglomerate portion, and the upper silts and fine

sand portion. Has relatively low

permeability; values range from 1 to 200 ft/day. Storage capacity correspondingly

low. In very minor part, a few beds of gravel and sand are sufficiently

clean that permeability is

moderately large; on the other hand, some beds of silty clay or clay are essentially impermeable.

Miocene and Pliocene Columbia River basalt series (10,000 ft thick) Basaltic lavas with interbedded sedimentary rocks, considerably

deformed. Underlie the unconsolidated sediments. Rocks are generally dense except for numerous shrink-age cracks, interflow scoria

zones, and interbedded

sediments. Permeability of rocks is small (e.g.,

0.002 to 9 ft/day) but transmissivity of a thick

section may be

considerable (70 to

700 ft 2/day)

10 0 5 1005

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-050 2.5-1 2.5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING

The information discussing the geology, seismol ogy, and geotechnical engineering is contained in a technical memorandum, TM-2143 , "Geology, Seismology, and Geotechnical Engineering Report." This report is incorporated by reference into the FSAR and as su ch is subject to the same controls as the FSAR.