ML14010A293
| ML14010A293 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/30/2013 |
| From: | Energy Northwest |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14010A476 | List:
|
| References | |
| GO2-13-174 | |
| Download: ML14010A293 (376) | |
Text
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 Generati on 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 Powe r Systems Criteria..............................................1.2-6 1.2.1.2.4 Radwaste System Criteria........................................................1.2-7 1.2.1.2.5 Auxiliary Sy stems Criteria......................................................1.2-7 1.2.1.2.6 Shielding and Access Control Criteria.........................................
1.2-7 1.2.1.2.7 Nuclear Safety Sy stems 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 E nvirons....................................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 Recirc ulation System...................................................1.2-14 1.2.2.4.4 Residual Heat Removal System.................................................1.2-14 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 Protect ion 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 Isola tion 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 Co re Spray System ........................................... 1.2-17 1.2.2.5.8.2 Automatic Depre ssurization System ......................................... 1.2-17 1.2.2.5.8.3 Low-Pressure Co re Spray System ........................................... 1.2-17 1.2.2.5.8.4 Low-Pressure C oolant Injection .............................................. 1.2-18 1.2.2.5.9 Primary C ontainment
............................................................. 1.2-18 1.2.2.5.9.1 F unctional Design
............................................................... 1.2-18 1.2.2.5.9.2 Drywell Coo ling System
....................................................... 1.2-18 1.2.2.5.9.3 Suppression P ool Cooling
..................................................... 1.2-19 1.2.2.5.9.4 Containm ent Spray
.............................................................. 1.2-19 1.2.2.5.9.5 Containment Atmo sphere 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 Restrict ors .............................................
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 - S econdary Containment .................................. 1.2-21 1.2.2.5.16 Reactor Building Ve ntilation 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 Po wer Supply System ......................................... 1.2-23 1.2.2.5.20 Standby Liqu id 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 COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 Chapter 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
TABLE OF CONTENTS (Continued)
Section Page LDCN-05-009 1-iii 1.2.2.5.23 Fuel Pool Coo ling and Cleanup System.......................................1.2-24 1.2.2.6 Power Conversion System.........................................................1.2-24 1.2.2.6.1 Turbine Generator.................................................................1.2-24 1.2.2.6.2 Main Steam System...............................................................1.2-24 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 Gla nd Seal System......................................................1.2-25 1.2.2.6.6 Steam Bypass System and Pressure Control System........................1.2-25 1.2.2.6.7 Circulating Water System........................................................1.
2-25 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 Pro cess Control and Instrumentation..........1.2-26 1.2.2.7.3 Nuclear System Process C ontrol and Instrumentation......................1.2-27 1.2.2.7.3.1 Reactor Manual Control System.............................................1.2-27 1.2.2.7.3.2 Recirculatio n Flow Control System..........................................1.2-27 1.2.2.7.3.3 Neutron Monitoring System...................................................1.2-27 1.2.2.7.3.4 Refu eling Interlocks............................................................1.2-27 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 Electr o-Hydraulic Control System..................................1.2-28 1.2.2.7.4.2 Feedwate r System Control....................................................1.2-28 1.2.2.8 Radioactive Waste Systems........................................................1.2-28 1.2.2.8.1 Liquid Radw aste System.........................................................1.
2-28 1.2.2.8.2 Solid Radwaste System...........................................................1.2-29 1.2.2.8.3 Gaseous Radw aste System.......................................................1.2-29 1.2.2.9 Radiation M onitoring and Control................................................1.2-30 1.2.2.9.1 Process Radia tion Monitoring...................................................1.
2-30 1.2.2.9.2 Area Radiat ion Monitors.........................................................1.2-30 1.2.2.9.3 Site Radiological E nvironmental 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 1.2.2.10 Shielding..............................................................................1.2-32 COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 Chapter 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
TABLE OF CONTENTS (Continued)
Section Page LDCN-04-037,09-019 1-iv 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 Handli ng System
............................................................ 1.2-32 1.2.2.11.3 Fuel Pool Cooling and Cleanup System ....................................... 1.2-32 1.2.2.12 Cooling Wa ter 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 San itary Drain Systems ................................... 1.2-33 1.2.2.12.6 Process Samp ling Systems
....................................................... 1.2-33 1.2.2.12.7 Condensate S upply System
...................................................... 1.2-34 1.2.2.12.8 Equipment and Floor Drainage Systems ...................................... 1.2-34 1.2.2.12.9 Compressed Ai r 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 Communicati ons Systems
........................................................ 1.2-37 1.2.2.12.13 Lighting Systems ..................................................................
1.2-37 1.2.2.12.14 Normal Auxiliary Alterna ting 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 RE GULATORY GUIDES ....................... 1.2-38 1.3 COMPARISON TABLES ................................................................ 1.3-1 1.3.1 COMPARISONS WITH SIMILA R FACILITY DESIGNS ..................... 1.3-1 1.3.1.1 Nuclear Steam Supply Syst em Design Charact eristics
........................ 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 Characteris tics .............................................
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 COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 Chapter 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
TABLE OF CONTENTS (Continued)
Section Page LDCN-07-011 1-v 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS.......................1.4-1 1.4.1 APPLICANT/O PERATOR...........................................................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 CONT RACTORS............................................................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 AREVA NP..........................................................................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 COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 Chapter 1
INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
LIST OF TABLES
Number Title Page 1-vii 1.2-1 Principal Regulati ons 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 Ma nagement 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 Co mpleted, Under Construction, or in Design by Ge neral Electric.............................................1.4-5
1.6-1 Topical Reports.................................................................1.6-3
COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT 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, Tu rbine Generator Building (Sheets 1 and 2) 1.2-3 General Arrangement
- Mezzanine Floor Plan, Tu rbine Generator Building (Sheets 1 and 2)
1.2-4 General Arrangement
- Operating Floor Plan, Tu rbine 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 a nd 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 Arra ngement - 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 Arrange ment - El. 437 ft 0 in., Radwaste Building
1.2-14 General Arrange ment - El. 467 ft 0 in. and Part ial Plans, Radwaste Building
1.2-15 General Arrange ment - El. 484 ft 0 in., El. 487 ft 0 in., and Partial Plans, Radwaste Building
COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT 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 Arrange ment - El. 501 ft 0 in., 507 ft 0 in., and 525 ft 0 in., Radwaste Building 1.2-17 General Arra ngement - Radwaste Building Sections (Sheets 1 and 2)
1.2-18 General Arrangeme nt - Service Building
1.2-19 General Arra ngement - Service Building Sections
1.2-20 General Arrangement - Standby Service Wate r Pump Houses Sections
1.2-21 General Arrangeme nt - Circulating Water Pump House Sections
1.2-22 General Arra ngement - Diesel Generator a nd Service Building Sections
1.2-23 General Arra ngement - Makeup Water Pump House, Plans and Sections
1.2-24 General Arra ngement - 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)
COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 LDCN-02-000 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 Generati ng 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 unde r 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 sy stem (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 P ittsburgh Des Moines Steel Co.
The vessel contains both a drywell and a suppression chambe r, 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 react or is 3486 MWt. The design power level limit is 3629 MWt. The net electr ical 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 opera tion on December 13, 1984.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 func tions are not always clear-cut and are sometimes overlapping, the functional classificati on 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 Gene ration Design Criteria
- a. The plant was designed so that it can be fabricated, erecte d, and operated to produce electric power in a safe and re liable manner. Plant design conforms to applicable codes and regula tions as stipulated in Table 1.2-1
- b. The plant is designed to produce steam fo r 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 ga ses released from fuel mate rial 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 ca n be manually controlled; COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 system and locate system malfunctions; and
- k. Interlocks or other au tomatic equipment are provide d as backup to procedural controls to avoid conditions requiring th e 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 material s to the environment is limited to the limits and guideline values of applicable federal regul ations 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 design ed so the discharge and offsite shipment of radioactive e ffluents can be made in accordance with applicable regulations;
- f. The design provides means by which pl ant 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 pr ovided and access control patte rns have been established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any mode of norma l plant operations;
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November1998 1.2-3 i. Those portions of the nuclear system that form part of the reactor coolant pressure boundary (RCP B) 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 fa ilure of active com ponents 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 system s 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 contro l of active components of nuclear safety systems and ESF from the control room;
- n. Nuclear safety systems and ESF are designed to perm it demonstration of their functional performa nce requirements;
- o. The design of nuclear safety systems and ESF includes allo wances for natural environmental disturbances such as ear thquakes, 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 circumst ances where offsite power sources are not available;
- r. Features of the plant that are essential to the mitigation of ac cident consequences are designed, fabricated, and erected to quality standards that reflect the importance of the safety action to be performed;
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-05-009 1.2-4 s. A primary containment has been provide d that completely en closes the reactor system, drywell, and suppr ession pool. The primary containment employs the pressure suppression concept;
- t. The primary containment is designed to re tain 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 containm ent integrity and leaktightness at periodic intervals;
- v. A secondary containment has been provided that completely encloses both the primary containment and fu el storage areas. Th e secondary containment includes the standby gas treatment (SG T) system for controlling release of radioactive materials leaking from the primary containment in the event of an accident and also has the cap ability for filtering radioactive materials directly from the primary containment atmo sphere during shutdown conditions;
- w. The secondary containment has been de signed to act as a ra dioactive material barrier, if required, when the primar y 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 remo ving energy from within the containment vessel as necessary to main tain 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 whenev er 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 (E CCS) 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 co re cooling over the complete range of postulated break sizes in th e RCPB and are redundant; COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-5 cc. Operation of the ECCS is initiated au tomatically when requi red, regardless of the availability of offsite power supp lies and the normal ge nerating system of the plant;
dd. The control room has been shielded against radiation and provided with a high efficiency filtration system so that continued occupanc y under accident conditions is possible; ee. In the event that the control room becomes inacce ssible, it is possible to bring the reactor from power range operation to cold shutdown co nditions 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 b ackup system has the capability to shut down the reactor from any normal oper ating 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 sh ielding 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 designe d 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 syst ems, is designed to retain integrity throughout any abnor mal 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;
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 sh utdown to design power and for any abnormal operational transient. The capac ity of such system s 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 norma l operational heat removal systems become inoperative. The cap acity of such systems is adequate to prevent fuel cladding damage. The reactor is capable of being au tomatically 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 wo rth 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 ther e is no tendency for di vergent oscillation of any operating characteristic, consideri ng the interaction of the nuclear system with other appropriate plant systems.
1.2.1.2.2 Power Conver sion System Criteria
Components of the power conversion system ha ve 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 pr oducts or radioactivity associated with the steam and condensate during normal operation are safely containe d 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 sa fe condition under all credible COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-7 circumstances. The power sources are adequate to accomplish all required engineered safety feature functions under postulated design basi s 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 dur ing normal operation within those limits specified in 10 CFR 20 and 10 CFR 50, Appendix I;
- b. The solid radwaste dispos al 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 th rough 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 shieldi ng and cooling for spent fuel
. Provision is made for maintaining the cleanliness of spent fuel cooli ng and shielding water;
- b. Other auxiliary systems, such as standby service water (SW), high pressure core spray (HPCS) SW, fire protecti on (FP), heating and ventilating,
communications, and lighting systems, are designed to func tion 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 radiati on doses within the limits of published regulations in any normal mode of plant operation; and
- b. The control room is shielded agai nst radiation and has a high efficiency filtration system, so that occupancy is possible under accident conditions and COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 syst ems 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 cha nges within design limits.
- 2. It is possible to manually c ontrol the reactor power level.
- 3. Control of the reactor is po ssible 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 co ntrol the reactor pressure throughout its operating range.
- 2. The turbine is able to respond auto matically 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 eq uipment is possible from a central location.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 ad equate to safely shut down the unit.
- 2. Protective relaying is used to detect and isolate fau lted equipment from the system with a minimum of dist urbance in the event of equipment failure.
- 3. Primary and secondary undervoltage relays ar e 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 po wer system diesel generators.
- 4. Standby power diesel gene rators' 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 ca n be operated from the main control room.
- 6. Metering for essential generators, tr ansformers, 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 cr iteria 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 app licable to individual systems or st ructures are disc ussed throughout the FSAR.
The plant shielding and radiation z one classification can be found in Table 1.2-2
. Chapter 12 provides further details.
COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT 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 southeas t area of the Department of Energy (DOE) Hanford Reservation in Benton Count y, Washington. The si te is approximately 3 miles west of the Columbia River at River Mile 352, approximately 12 miles north of the City of Richland, 18 miles north west 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 charac teristics of the site which make it well-suited for operation of the plant include: favorable geographical, geological
, and seismological characteristics; ad equate water supply; id eal climatological charact eristics; 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 world's first nuclear pr oduction reactors. Since 1943, nine plutonium production reactor s and a number of test reactor s have been constructed and operated at the Hanford Site.
1.2.2.1.2.4 Meteorology
. The climate around CGS is basically continenta l 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 discharg e boundary for the aquifer. The surface soils at Hanford are sufficiently pe rmeable 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 la vas and intercalated sediments of the Columbia River Group at the ba se, (b) Plio-Pleisto cene 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.
COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 LDCN-04-052 1.2-11 1.2.2.1.2.7 Seismology
. The CGS site is situated in an area charac terized by low seismicity and widely scattered epicenters. See Section 2.5 for additional information.
1.2.2.1.3 Design Basis Depe nding on Site Environs
- a. Offgas System
An offgas (OG) system consisting of hold-up pi ping, charcoal adsorbers, and an elevated release is provided for the contro lled 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 recycl ed, 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 to rnado 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 upstr eam dams due to seismic forces, 2. High flow in the Columbia River, and
- 3. Wind and wave action.
COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT 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 bu ilding that houses the standby diesel generators, diesel fuel o il (DO) storage tanks, and associated controls and instrumentation;
- e. Circulating water pump house (Wind River Building) - a structure housing the main circulating water (C W) 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 ulti mate heat sink (UHS);
- h. Makeup water pump house - a structure that hous es the cooling tower makeup (TMU) water pumps;
- i. General service building (Yakima Building) - a struct ure 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).
COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT 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 Electric's (GE) pi ping 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 di rect use in the steam turbine.
A heat balance showing the major parameters of the nuclear system for the rate d 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 a ssemblies. The control rods are positioned by individual control rod drives (CRDs).
Each fuel assembly has several fuel rods with gadolinia (Gd 2O3) mixed in solid solution with UO2. The Gd 2O3 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 desi gn 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 th e control rods are not su sceptible to distortion and have an average life exp ectancy many times the residence time of the fuel loading.
1.2.2.4.2 Reactor Vessel and Internals The reactor vessel contains the co re and supporting structures; th e steam separato rs 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 reci rculation (RRC) lines, RFW li nes, CRD and in-core nuclear instrument housings, HPCS and LPCS lines, re sidual heat removal (RHR) lines, SLC line, COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT 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 fabric ated 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 reacto r 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 vess el but inside the primary containment. Each external loop contains a mechanical pump, two motor-operated mainte nance valves, and one flow control valve which is mechanically blocked full open. The two motor-operated valves are used as pump suction and pu mp 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 th e jet pumps, which contai n 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 vessel's inner wall. Any recirculation lin e 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 He at 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 th e reactor vessel, following a LOCA, rapidly enough to reflood the core and maintain fuel cla dding below the fragmentation temperature independent of other core cooling systems. This is further discussed in Section 1.2.2.5.8
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-05-009 1.2-15 c. Removes heat from the primary cont ainment, following a LOCA, to limit the increase in primary containment pressure. This is accomplished by cooling and recirculating the suppression pool water (containment cooli ng) 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 impur ities 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 ch anges, fill-up rate of drain sumps, and fission-product concen tration inside the primary containment. Large leaks are also detected by changes in reactor water level and ch anges in flow rate s 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 s hutdown (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 COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT 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 monitori ng system qualify as a nuc lear safety system, those that provide high neutron flux signals to the RPS do. Th e intermediate range monitors (IRMs) and average power range monitors (APR Ms), which monitor neut ron 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 provi de 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 housi ng supports are located underneath th e 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 nucl ear 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 co re 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 consisti ng of safety/relief valves (SRVs) mounted on the MS lines is provided to prevent excessive pr essure inside the nuclear syst em following either abnormal operational transients or accidents.
COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December2005 LDCN-04-052 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 ope rates automatically in time and with sufficient coolant flow to maintain adequate water level in the reactor vessel.
1.2.2.5.8 Emergency Co re 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 in jection (LPCI), an operating mode of the RHR system.
1.2.2.5.8.1 High-Pressu re Core Spray System. The HPCS system pr ovides and maintains an adequate coolant inventory inside the reactor vessel to maintain fuel cladding temperatures below the fragmentation temperat ure in the event of breaks in the RCPB. The system is initiated by either high pressure in the drywell or low water level in th e vessel. It operates independently of all other systems over the entir e 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. Th e depressurization provided by the system enables th e 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 c onditions indicate that the HPCS sy stem is not delivering sufficient cooling water to the reactor vessel to maintain the water level above a pr eselected 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-Pressu re Core Spray System. The LPCS system cons ists 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. Th is system provides the capability to cool the fuel by spraying water into each fuel channel. The LPCS loop COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-05-009 1.2-18 functioning in conjunc tion 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 ope rating mode of the RHR system, but is discussed here because the LPCI mode acts as an engineered safety feature in conjunction with other ECCS. Th e 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 deliver ed to the core only after reactor vessel pressure is reduced. The LPCI operation provides th e capability of core reflooding, following a LOCA
, in time to maintain the fuel cladding below the prescribed te mperature limit.
1.2.2.5.9 Primary Containment
1.2.2.5.9.1 F unctional Design
. The primary containment is part of the overall containment system which provides th e 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 wh ich stores a large volume of water, a connecting submerged vent system between the drywell and water pool, isolation valves, contai nment 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 dr ywell air space. The resulting increase of drywell pressure would then force a mixture of air, steam, and water thr ough the vents into the pool of water which is stored in the suppressi on pool, resulting in a rapi d pressure reduction in the drywell. Air which is transferred to th e suppression chamber, pr essurizes 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-handli ng units to maintain the required ambient temperature. Air is distributed through ductwork and/or up th rough the annular space between the reactor vessel insulation and th e 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 th e 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 esse ntial buses to ensure conti nuous operation to satisfy the power-generation design objective.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-05-009 1.2-19 The drywell cooling system is designed to opera te during offsite power loss. Control switches
for operating the equipment are loca ted 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 Containmen t Atmosphere Control. In the event of a LOCA, hydrogen and oxygen will be generated in the reactor. C ontainment atmosphere control is provided by inerted containment, containment atmosphere mixing, and hydrogen an d 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 act ivated 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 ons et 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 w ith redundant isola tion capabilities.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 c ontainment and offer a pot ential 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 isol ation consideration.
Automatic MSIVs are provided in each MS line. Each is pow ered 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 spikin g) by isolating the RCPB in case of a rapid depressuriza tion 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 th e containment. The monitors are designed to detect a gross release of fission products from th e fuel. On detection of high radiation, the trip signals genera ted by the monitors are used to initiate a closure to the reactor water sample valves, mechanical vacuum pum p 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 re dundant systems. Each system consists of a pump and piping supplying the associated RH R system heat excha nger, standby diesel generator, essential heating, ventilating, and air conditioning (HVAC) coolers, RHR pump seal
coolers, SW motor bearing coolers, and sample coolers with sa fety grade cooling water from COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 LO CA 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 fo r the shutdown function of the RHR system.
The SW is available to the she ll side of the fuel pool cooling and clean up (FPC) system heat exchangers in the event that the normal coo ling water supply from th e 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 Buildi ng - Secondary Containment
The reactor building completely surrounds the primary containment. The building provides secondary containment when the primary containmen t is closed and in se rvice, 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 react or 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 syst em consists of a number of radiation monitors arranged to monitor the activity level of th e ventilation exhaust from the reactor building and primary containment. Up on detection of high ra diation, the reactor building is automatically isolated and the SGT system is started.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-04-050 1.2-22 1.2.2.5.17 Standby Ga s 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 cap able of exchanging the total reactor building volume once in a 24 hr period.
Each filter unit contains electric heaters, a prefilter, 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 s lightly negative internal building pr essure and can process all gaseous effluent prior to its discharg e from the reactor building.
All equipment is connected to the essential bus es 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 generato r sets, switchgear, and associated distribution system equipment and auxiliaries.
These diesel generator sets ar e associated with redundant (D ivisions 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 norm al and LOCA conditions, in the event that both the offsite and the normal auxiliary power sources are unavailable to supply plant loads. Since load distri bution is such that redundant auxiliary systems are separated by division, safe shutdown can be achieved with only one of th e two diesel gene rators operating.
The standby ac power supply system diesel gene rators and associated equipment are designed to Class 1E standards and are lo cated within Seismic Category I st ructures. Equipment of each division is separated so that fa ilure 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
.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-05-009 1.2-23 1.2.2.5.19 Direct Current Power Supply System
The dc power supply system c onsists of station batteries, battery chargers, distribution equipment, and re lated 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 start up range and power range neutron monitoring systems.
The primary power sources for the system are the dc output stat ion battery chargers. Station batteries associated with each charger operate in a "float-charge" configuration to ensure maintaining the batteries in a fu lly 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 divi ded into electrically and physically independent divisions. Each battery, together with its independent battery charger, is associated with one of the se gregated divisions. The batteries and their associated chargers are located in separate rooms.
The ampere-hour capacity of each battery is capable of suppl ying 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 Liqu id Control System
Although not intended to provide prompt reactor shutdown, as the control rods are, the Standby Liquid Control (SLC) sy stem provides a redundant, inde pendent, and alternate method to bring the nuclear fission reacti on to subcriticality and to maintain a subcritical condition as the reactor cools. The system makes possible an orderly and sa fe shutdown in the event that not enough control ro ds can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract th e positive reactivity effect from rated power to the cold, cl ean 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 gase ous 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-05-009 1.2-24 1.2.2.5.22 Main Steam Line Isolation Va lve Leakage Control System (Deactivated)
The main steam line isolation valve leakage control (MSLC) sy stem 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 be tween 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 st ored spent fuel and maintains specified water temperature, purity
, clarity, and level. This pr events boiling of the pool water and controls the buildup of excessive radioactive materials in the cooling water, thereby minimizing potential radia tion exposure to plant personnel.
The cooling portion of the system is designed to Seismi c Category I requirement s and may be isolated from the Seismic Category II cleanup portion of the system by au tomatic Seismic Category I isolation valves which actuate on low-fuel pool wa ter level. Normally the RCC system furnishes non-safety grade cooling water to the FPC system. If requi red, safety grade coo ling and makeup water is available to the FPC system from the SW system.
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-pr essure turbine and three double-flow low-pressure turbines), reheat unit with an electrohydraulic governor for normal operation. The turbine generator is provid ed with an emergency trip system for turbine overspeed. The rating of the tu rbine generator is 1,173,046 kW.
The generator is a direct-driven, thr ee-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 hyd rogen 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 MS IVs to the main turbine stop valves. The use of four main steam lines permits testing of th e 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 in clude drains and parts of the turbine bypass system.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-06-057 1.2-25 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, a nd condensate pump suction vents.
1.2.2.6.4 Main Condens er 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 ope ration. The system consists of two 100%-capacity, twin-element first stage and si ngle-element sec ond stage steam jet air ejector units complete with intercondensers for normal plant operation and a mechanical vacuum pump for use during startup. Discha rge 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 pipi ng, valves, and instrumentation.
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 pe rmitted to pass to the turbine ge nerator. 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 co nstant reactor pressure during normal plant operation.
1.2.2.6.7 Circula ting Water System The CW system provides the condenser with a c ontinuous supply of cool ing 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 Columb ia River. The makeup water replaces the water lost by evaporation, drift, and blowdown.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-26 1.2.2.6.8 Condensate a nd Feedwater System The condensate and feedwater system pumps co ndensate from the condenser hotwell to the RPV. Condensate is pumped by three main condensate (COND) pumps th rough the gland seal steam condenser, the steam jet air ejector condensers, and the offg as condenser. After leaving the offgas condenser, the condensate is pumped throu gh a full-flow condensate filter-demineralizer syst em. 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 feedwa ter heating and then flows to the RPV.
Feedwater flow is controlled by varying the speed of th e steam-driven turbine.
1.2.2.6.9 Condensate Filter-Demineralizer System
The full-flow condensate filter-d emineralizer system with instrumentatio n 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 ma in generator directly connected to a main power transformer through an isolated phase electri cal bus duct. The main power transformer steps up the output of the 25-kV generator to a nominal 500-kV transmission system voltage.
The output of the main power transformer is connected to a 500-kV switchyard consisting of circuit breakers, disconnect switches, buses, and associated e quipment arranged in a ring bus configuration.
A 230-kV offsite supply is provi ded to a separate startup a uxiliary transformer to supply maximum startup, operating and shutdown load requirements for a normal plant auxiliary loads and for safety loads. In add ition, a separate 115-kV offsite supply serves a backup auxiliary transformer with sufficient capac ity to provide the power requirements of plant safe shutdown loads.
1.2.2.7.2 Electrical Power Systems Process Control an d Instrumentation Main generator electrical controls are located in the main control room. These include main generator circuit breaker controls, synchronizing equipment, and ge nerator excitation and voltage control equipment. Inst rumentation is also provided in the main control room for the main generator connections and equipment. This includes indi cating instruments for voltage, current, kW, MVAR, and frequency. Recordi ng instruments are provided for generator MW output and main bus volta ge. Kilowatt-hour mete rs are provided for ma in generator outputs COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-27 and for auxiliary power system loads. Instru mentation is provided fo r monitoring generator and transformer temperatures.
Other types of monitoring instrumentation are provided as required to ensure proper opera tion of equipment. Circuit br eaker controls, metering, and indication for the auxiliary power system ar e 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 include s 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 Proces s 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 va lves 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 certa in conditions as a backup to procedural controls.
1.2.2.7.3.2 Recirculat ion 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.
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 fl ux, 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 Refu eling 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 cr iticality during refueling operati ons. 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-06-057 1.2-28 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 tr ansmit information that can be used to assess conditions existing inside the reactor vess el and the physical condition of the vessel itself. This instrumentation monitors reactor vessel pressure, water level, coolant temperature, reactor core differ ential pressure, coolant flow ra tes, 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 reactiv ity 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 Electr o-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 tu rbine valves and bypass valves.
The turbine generator speed/load controls can initiate rapid cl osure of the turbine control (governor) valves and rapid openi ng of the turbine bypass valves to prevent turbine overspeed on a generator electric load loss.
1.2.2.7.4.2 Feedwa ter 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, reac tor 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, shutdow n, and low plant load conditions, the steam tu rbine-driven feedwater pumps are run at constant speed, a nd the feedwater contro l 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 Ra dwaste System
This system collects, treats, stor es, 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 COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-29 the quality and quantity of material s determined to be present. Processed liquid wastes may be returned to the condensate system or discharged to the circula ting 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 belo w the limits of 10 CFR Part 20.
Equipment is selected, arranged, and shielded to permit operati on, inspection, and maintenance with minimum personnel exposure. For example, tanks and processing equipment which contain significant radiation sources are located behind sh ielding, and sumps, pumps
, instruments, and valves are located in c ontrolled 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 abnorm al 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 st eel 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.
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 conde nser are the major source of ga seous radioactive waste. The treatment of these gases includes volume reduction through a cataly tic hydrogen-oxygen recombiner, water vapor removal through a cond enser, decay of shor t-lived radioisotopes through a holdup line, further condensation, filtration, adso rption of isotopes on activated charcoal beds, further filtration through high efficiency filters, and final release.
Continuous radiation monitors are provided which indicate radioac tive release from the reactor and from the charcoal absorbers. The radiati on 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 gla nd seal steam condenser are not treated prior to release.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-30 The design of the OG system is such that th e annual exposure to a ny offsite person during normal operation from gaseous sources wi ll be ALARA and less than 10 CFR 20.
1.2.2.9 Radiation M onitoring 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 pr ocess liquids and gases or for process system malfunctions. All effluents from the plant which are potentially radioactive are monitore
- d. 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,
- f. Reactor building ventilation exhaust ple num 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 th roughout the plant buildings to ensure that plant personnel will not be ina dvertently exposed to high radiation doses.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-31 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 eight different media.
Ambient radiation dose will be monitored us ing thermoluminescen t dosimeters (TLDs).
Airborne particulates are me asured by filtering known quantities of air and analyzing the filtered material. Radioiodine in the air is meas ured in the same way except it is adsorbed onto a charcoal cartridge rather than being filtered.
Water is sampled at the plant intake, from the plant discharge, in the ri ver below the plant, and at the nearest downstream municipal water supply. Groundwater in th e vicinity is also sampled.
The radiation monitoring program includes sampling of garden produce where available in the vicinity of the site, the collection of river se diment samples from a bove and below the plant discharge point, the collection of fish samples from the Columb ia River and the Snake River, and the collection of milk samples at four or more locations near the site.
The details of this monitoring program are given in Section 5.0 of the Offsite Dose Calculation
Manual (ODCM).
1.2.2.9.4 Liquid Radwas te System Control
Liquid wastes to be discharged are handled on a batch basis with prot ection against accidental discharge provided by procedural controls. Instrumentation with alar ms 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 Radwas te System Control
The solid radwaste system collects, treats, and stores so lid radioactive wastes for offsite shipment. Wastes are handled on a batch basis. Radiation leve ls of the various batches are monitored by the operator.
1.2.2.9.6 Gaseous Radwaste System Control
Gaseous radwastes are discharged through a reac tor building elevated re lease point. Radiation levels of the release are continuously monitored and recorded.
Isolation of the main condenser offgas is automatically initiated prior to rel ease should the activity of the offgas exceed discharge limits.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-32 1.2.2.10 Shielding The shielding in the plant is designed to mini mize exposure of plant pe rsonnel 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 r acks are designed to prevent inadvertent criticality and load buckling. Sufficient coolant and shielding are maintained to prevent overheating and excessive personnel exposure, respectivel
- y. 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 cr anes, 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 require d, safety grade cooling and make up water from the SW system is available to the system by remote-manual operati on of redundant Seismic Category I valves to provide long-term cooling and prevent fuel pool boiloff which might result in unacceptable building environmental conditions.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 system s 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 equipmen t located within the plant.
1.2.2.12.3 Ultimate Heat Sink
Two spray ponds that serve as the UHS conserva tively 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 demine ralizers and the DW system.
The DW system is designed to provide demine ralized water to the CS Ts for plant makeup and demineralized water for other plant opera ting requirements.
1.2.2.12.5 Potable Water and Sanitary Drain Systems
The plant potable water (PW) sy stem 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 se ries with lined facu ltative 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 DOE's 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. Repres entative liquid and COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 1.2-34 gas samples are taken automa tically and/or manually dur ing 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 head er, to the CRD pumps via a second header, and to the RHR, RCIC, and HPCS systems and conde nsate supply and condensate fi lter/demineralizer backwash pumps via a third header. The condensate supply pumps deliver condens ate 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 condensat e 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 drai nage systems which do not carry radioactive waste.
All drains in the reactor buildi ng and radwaste building are cons idered radioac tive. 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 wate r drainage system. The storm water drainage system is normally nonradioactiv e, however some accumulati on 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, o il-free air to station instrumentation and controls a nd 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 fo r maintenance throughout the station and breathing air at selected locations.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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, th e critical switchgear area, the cable spreading room HVAC systems, the SW pump room heat removal systems, the reactor building
emergency pump and critical electric equipmen t area cooling systems, and the ventilation system for the standby diesel generators are de signed to operate under a ll 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 pot ential contamination.
Three separate and redundant HVAC systems service the ma in 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 pr ovide 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 before release to the atmosphere.
The SW pump room heat removal systems consis t 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 ven tilation system consists of five fan coil units and nine recirculation fans. During normal operation, a minimu m three out of five fan coil units are operating.
Ventilation for the reactor building is provided by a once-through ven tilation system based on the use of evaporative coolers. The system incorporates the necessary isolation valves to COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 a ppropriately throughout the plant and fire alarms are annunciate d in the main control room.
The FP system provides a reliable water di stribution system for extinguishing fires. Two motor-driven fire pumps are used for norma l service, with a dies el-engine-driven fire pump as a backup. A second diesel-driven fire pump with a sepa rate 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 a nd most other areas c ontaining 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 c ontrol room power generation control complex (PGCC) subfloor.
Wet pipe sprinklers protect the turbine/generator beari ngs and other miscellaneous areas. Preaction sprinkler systems protect diesel generators, day tank/transfer pump rooms, and areas with high concentra tions of electrical cables.
Manual suppression includes:
Fire hydrants spaced around the yard fire main loop.
Fire hose stations locate d throughout the plant.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 relia ble communication inside and outside the plant and from the plant to local fi re protection and law enforcement authorities.
The system utilizes a public address and building wide alarm system, a public telephone system, a private digital te lephone system, a sound powere d telephone system, a radio communication system, and an au tomatic 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 li ghting, dc lighting, and battery-pack emergency lighting. Lighting in tensities are designed to provide indoor and outdoor illumination consistent with the Ju ly 1974 Illumination Engineering Society recommendations, and meet or exceed Occ upational 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 auxilia ry 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 120/20 8-V non-Class 1E distribution system.
The normal ac auxiliary transformers provide power to all plant auxiliari es and comprise the normal plant ac power source when the main generator is ope rating. One of the normal auxiliary transformers is a dua l secondary type with both sec ondary windings stepping down the generator voltage to 4.16 kV for supply to 4.16-kV non-Cl ass 1E switchgear buses. The other normal auxiliary transformer steps down th e 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 switchge ar. The ESF portions of the 480-V distribution system are supplied via the 4.16-kV Class 1E switchgear, and therefore ar e 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-09-019 1.2-38 1.2.2.12.15 Diesel Gene rator Fuel-Oil Storage and Transfer System
The diesel fuel oil storage and transfer system consists of separate, independent diesel oil supply subsystems servi ng each of two emergency diesel ge nerators and the HPCS diesel
generator. Each full capacity s ubsystem 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) syst em normally operates only when the heating steam evaporators are inoperative during plant shut down. The system then supplie s steam to HVAC systems for air and water space heating and for humidificati on 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 meteri ng pump, deaerator and boiler f eed pumps, condensate return tank pumps, steam supply and condensate return pi ping and valves, and a ssociated instruments and controls.
1.2.3 COMPLIANCE WITH NRC REGULATORY GUIDES
The CGS conformance to the NRC regulat ory guides is documented in Section 1.8 and in appropriate sections of this FSAR.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.2-39 Table 1.2-1 Principal Regulations and Code s Followed in Plant Design Number Tit le 10 CFR series Code of Fede ral Regulations, principally:
10 CFR 20 Standards for Prot ection Against Radiation 10 CFR 50 Licensing of Producti on 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 Po wer Generating St ation Protection Systems IEEE-308 IEEE Criteria fo r Class IE Electrical Systems for Nuclear Power Generating Stations ASME B&PV ASME Boiler a nd Pressure Vessel Code: Section III Nuclear ComponentsSection VIII Pressure VesselsSection XI Inservice Inspection
AEC Press Release IN-817 Tentative Regulatory Supplem entary Criteria for ASME Code-Constructed Pressure Vessels ANSI-B31.1.0 ANSI Standard Code for Pressure Piping, Power Piping
NOTE: Additional codes and regul ations applying to specific areas of system design are referenced in discussions of individual systems.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 w hole body dose is not e xpected to exceed 1.25 rem per calendar quarter.
- 3. Areas having dose rates in excess of 100 mr em/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 controlle d check points.
- 7. Controlled and limited access areas are identified by warning signs.
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Draw. No.
Rev.Figure 01.2-26Amendment 60 December 2009 Flow Diagram Legend,Symbols and Abbreviations M501Columbia Generating Station Final Safety Analysis ReportDraw. No.Rev.FigureForm No. 960690FH 51 HSSFHSV HTSystem Acronyms 950021.43 1.2-27System Acronyms Diesel OilDiesel Building Outside Air Diesel Building Return Air Diesel (Engine) Starting Air Demineralized Water
Equipment Drains Radioactive
Electrical Maintenance Equipment Emergency Offsite Facility
Environmental Rad. Monitoring Exhaust Steam (Turbine)
Facilities Generic Equipment
Floor Drain
Floor Drain Radioactive
Fuel Oil Fire Protection
Fuel Pool Cooling Filtered Water Guard House Exhaust Air
Guard House Fire Protection
Main Guard House Guard House Mixed Air Guard House Outside Air Guard House Water Hot Potable Guard House Return Air
Glycol Hydrogen (Turbine Generator)
Heating Steam Condensate Heater Drain Heating Hot Water
Health Physics
Heating Steam Hydrogen Storage and Supply Facility
Heating Steam VentHeat TracingHeater VentHydrogen Water Chemistry
RRC Hydraulic Control
ISO Phase Bus Duct Cooling
Instrument Rack
Chemical Feed
Leak Detection
Laboratory Equipment (Permanent Plant)
Laundry Facility
Low Pressure Core Spray
Loose Parts Detection System
Local Power Range Monitor
Miscellaneous Drain
Mechanical Maintenance Equipment
Master Equipment List
Meteorological
Mobile Laundry Facility
Main Steam (Nuclear)
Machine Shop Equipment Spent Fuel Storage Standby Gas Treatment
Service Building Heating Condensate Service Building Heating Hot Water
Salinity Monitoring Service Building Mixed Air
Seal Oil Suppression Pool Temp Monitoring Service Building Potable Hot Water Service Building Return Air
Source Range Monitoring
Sealing Steam Standby Service Water Solid Waste Service Water Chemical Feed Transient Data Acquisition System Turbine Building Exhaust Air Turbine Generator Traversing Incore Probe Tower Makeup Water Turbine (Lube) Oil Turbine Building Outside Air Turbine Bldg. Potable Hot Water Turbine Building Return Air Technical Support Center Plant Service Water Vessel (Sect. 8, Non Power Block)
Variable Speed Drive Bldg. Mixed Air Radwaste Building Chilled Water Radwaste Building Exhaust Air
Radwaste Heating Condensate Radwaste Building Mixed Air Washington Nuclear Plant 2 (Columbia Generating Station)
Radwaste Building Outside Air Radwaste Bldg. Potable Hot Water Radwaste Building Return Air
Radwaste Building Refrigeration
Wide Range Monitoring
Chemical Feed System MSLC MT MW MWR NSSE NSSS OFEA OFMA OFOA OFRA OG OL PDIS PEA PI PL PMA POA PPC PRA PRM PS PSD PSR PVMS PVR PWC PWH PWR RBM RCC RCICRDREARFT RFW RHR ROA RPIS RPS RPWH RRA RRC RSE RWCU RWMSSASAT SCH SCI SCW SEA SEC SEISSFS SGT SHCO SHHW SLC SM SMA SO SPTM SPWH SRA SRM SS SW SWA SWCF TDAS TEA TG TIP TMU TO TOA TPWH TRA TSC TSW VES VRMA WCH WEA WHCO WMA WNP2 WOA WPWH WRA WRE WRM ZINCMain Steam Leakage Control (Deactivated)
Material Transport Miscellaneous Waste Miscellaneous Waste Radioactive
Nuclear System Servicing Equipment
Nuclear Steam Supply System Offsite Facility Exhaust Air Offsite Facility Mixed Air Offsite Facility Outside Air Offsite Facility Recirculation AirOff Gas Obstruction Lighting
Plant Data Information System Pumphouse Exhaust Air
Process Instrumentation
Plant Equipment Pumphouse Mixed Air Pumphouse Outside Air
Plant Process Computer Pumphouse Return Air
Process Radiation Monitoring
Process Sampling
Plant Sanitary Drain
Process Sampling Radioactive Plant Vibration Monitoring System Process Vents Radioactive Potable Cold Water Potable Hot Water Process Waste Radioactive
Rod Block Monitor Reactor Closed Cooling Water
Reactor Core Isolation Cooling Roof DrainReactor Building Exhaust AirReactor Feedwater Turbine
Reactor Feedwater
Residual Heat Removal Reactor Building Outside Air
Rod Position Indicator System
Reactor Protection System Reactor Building Potable Hot Water Reactor Building Return Air
Reactor Recirculation Reactor Service EquipmentReactor Water Cleanup Rod Worth Minimizer SamplingService AirSulfuric Acid Treatment Service Building Chilled Water
Supervisory Control
Stator Cooling W aterService Building Exhaust AirPlant Security
Seismic Monitoring SystemAlternate Access Point Bldg. and AppurtenancesTech. Support Cntr. Exhaust AirTech. Support Cntr. Mixed Air Annunciators Tech. Support Cntr. Outside AirAverage Power Range Monitors Tech. Support Cntr. Potable Hot Water Air Removal Tech. Support Cntr. Return AirTech. Support Cntr. Refrig. Equipment
Alternate Rod Insertion
Area Radiation Monitoring
Auxiliary Steam Backwash Air Breathing Air Supply
Boiler Chemical Feed
Cond. Blowdown or Rad. Boards
Bleed (Extraction) Steam Containment Atmosphere Control (Deactivated)
Control Air System Circ. Water Blowdown Control Room Chilled Water
Containment Exhaust Purge
Chemical Feed
Chemistry Equipment Containment Instrument Air Cooling Jacket Water
Chlorine Containment Monitoring System
Containment Nitrogen Condenser Drains & Vents
Condensate (Auxiliary)
Carbon Dioxide Communications Condensate (Nuclear)
Condensate Demineralizer Containment Recirculating Air
Control Rod Drive
Containment Supply Purge Cooling Tower Electrical Bldg. Mixed AirContainment Vacuum Breakers Circulating Water
CRD Decontamination Diesel Cooling Water
Diesel Exhaust (Engine)
Diesel Building Exhaust Air
Digital Electro-hydraulic Control
Diesel Generator
Diesel Lube Oil Diesel Building Mixed Air AAP AEA AMA ANN AOA APRM APWH AR ARA ARE ARI ARM AS BA BAS BCF BD BS CAC CAS CBD CCH CEP CF CHEM CIA CJW CL CMS CN CND CO CO2COMM COND CP CPR CRA CRD CSP CTMA CVB CW DCN DCW DE DEA DEH DG DLO DMADO DOA DRA DSA DW EDR ELEC EOF ERM ES FAC FD FDR FO FP FPC FW GEA GFP GH GMA GOA GPWH GRA GY H2HCO HD HHW HP HPCSHSHVHWC HY IBD IR IRM IRON LD LE LF LPCS LPDS LPRM MD MECH MEL MET MLF MS MSHColumbia Generating Station Final Safety Analysis ReportDraw. No.Rev.FigureAmendment 60 December 2009 Form No. 960690FH LDCN-08-000 Amendment 54 April 2000Equipment Acronyms 950021.42 1.2-28.1Form No. 960690 FigureDraw. No.Rev.Equipment AcronymsAAAudio AlarmACAir Conditioning UnitACCAccumulator ACMAcoustic Monitor/SensorADAir Damper AHAir Handling Unit
AIAir 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 CCompressorC024 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 Chiller CRACrane CRBControl Rod Blade CRMControl Module CRSConductivity Recorder Switch CRTTerminal Display Screen CSConductivity Switch CSKShield Transfer CaskCTCurrent Transformer/Cooling Tower CUCondensing Unit
DDamper (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 Recorder DRVEDrive Mechanism For CRD DSDensity Switch DT Dens Trans Or Drive Turbine DTISDiff Temp Indicating Switch DTRSDiff T emp Recording SwitchDTSDiff 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
EIPower 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
FFilterF/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
FIFlow 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.
GATEGate GCALAGS Calibrator GENGenerator GOVGovernor GVTGravity Ventilator
HHeaterH2EHydrogen Element H2IHydrogen Indicator H2ISH2 Indicating Switch/Monitor H2ITHydrogen Ind Transmitter H2RHydrogen Recorder H2THydrogen TransmitterHASHigh Amplitude SelectorHCHeating Coil HCUHydraulic Control Unit HFHEPA Filter HMHour Meter HOHydraulic Operator HOIHoist 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
ILIndicating LightIMDInductive Motor Drive INInverter INDInductor INDXIndexer IOSCurrent Operated Switch IRInstrument Rack
ISIntake SilencerISOLIsolator, Isolation Device ITDCurrent Transducer
IXIon ExchangerJBJunction Box JPJet Pump KBDComputer Keyboard (Security)
LLubricatorLALightning Arrestor LAGDynamic Compensator LASLow Amplitude Selector LCLevel Controller LCRMLog Count Rate Meter LCVLevel Control Valve LELevel Element LFLighting Fixture LGLevel Glass
LILevel Indicator LICLevel Indicating ControllerLISLevel Indicating Switch LITSLevel Indic Trans SwitchLMSLimit Switch LMTRV/I Signal Limiter LNRLinear Reactor LOCLube Oil Conditioner LPLighting Panel LPW24 Volt Lambda Power Supply LRLevel 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 Selector LWRUnknown Equipment Type ?LWSLow Differential Pressure
MMotorM/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 MXMixerColumbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Equipment Acronyms (Con't) 950021.44 1.2-28.2Form No. 960690 FigureDraw. No.Rev.Equipment Acronyms (con't)
MZNR O/M O2/H2O2EO2H2RO2IO2ROS OSC OZG P
P/B P/E P/I P/P PA PBU PC PCV PDM PDP PE PH PHB PHC PHE PHEC PHED PHIC PHIT PHITS PHR PHT PI PIC PICS PIS PL PLC PNL POC POE POI POIC POS POT POTR POV PP PR PRN PROG PRTM PRV PS PT PTA PTD PTZM PUI PUIT PUS QCC QDC QHM QSV R
R/I RA RAD RC RCM RD RDCC RDD REMultizone Air Conditioner Neutral Grounding Resistor
Input/Out Module
Oxygen /Hydrogen 2Oxygen Element 2Oxygen/Hydrogen Recorder 2Oxygen Indicator 2Oxygen Recorder 2Oil Separator
Oscillograph
Ozone Generator
Pump Push Button
Pneumatic/Electric Converter
Pressure/Current Converter
Pressure Inverter
Pre-Amps Seismic Playback Unit
Pressure Controller Pressure Control Valve
Power Distribution Module
Power Distribution Panel
Pressure Element Ph Ind Transmitter Recorder
Pneumatic Hydraulic Booster
Ph Controller
Ph Element
Photoelectric Controller
Photoelectric Detector
Ph Indicating Controller Ph Indicating TransmitterPh Indicating Transmitter Switch Ph Recorder Ph Transmitter
Pressure Indicator
Press Indicating Controller
Press Indicating Controller and Switch
Pressure Indicating Switch
Programmable Logic Card
Programmable Logic Controller
Panel Disc Position Signal Conv
Position Indication Element
Position Indicator
Position Indicating Controller
Position Switch Position Transmitter Potentiometer "CL.1E Only" Pilot Operated Pop Off Valve
Power Panel
Pressure Recorder
Line Printer
Programmer Programmable Timer Pressure Reg. Valve
Pressure Switch Poten. Xmfer Or Press. Transm.
Barometric Pressure Amplifier Pressure Transducer Pan Tilt Zoom Monitor
Purity Indicator Purity Indicator Transmitter
Purity Switch
Quick Couple Connection
Quick Disconnect Run Time Meter Quick Acting Solenoid Valve
Reservoir
Resistance/Current Converter Radiation Amplifier
Radiation Mon. Control Board
Radiation Controller
Respirator Cleaning Module
Rupture Disc
Rod Drive Control Cabinet
Rod Detector Display
Radiation Element RECT REL RES RF RFM RG RI RIS RLY RM RMC RMS RO ROD RPIS RPV RR RRM RSA RSCC RSDP RSM RSMD RSR RSRT RST RT RTM RV RVT S
SC SCAN SCL SCR SE SEW SF SH SHRED SI SIOA SL SM SMA SMD SNB SOL SP SPC SPS SPV SPVD SQRT SR SRU SS SSW ST SUH SUM SUMP SV SYNC T
T/SS TA TAPE TAS TB TBE TBIT TBR TBS TBT TC TCV TD TDS TE TE/MERectifier
Relay Resistor Refrigeration Machine(OG)
Radio Frequency Monitor
Regulator
Radiation Indicator
Radiation Indicating Switch
Relay Radiation Monitor
Remote Manual Controller
Remote Manual Switch
Restricting Orifice
Rod Position and Info Sys.
Radiation Recorder
Refrigerant Recovery Machine Response Spectrum Annunciator
Rod Sequence and Control Cab
Rod Sequence Display Panel
Radiation Sampler
Rod Select Module
Response Spectrum Recorder RSR Transducer for RSA Resin TrapRadiation Transmitter Run Time Meter Relief Value Roof Ventilator Electronic Trip Unit Speed or Seismic Controller
Scanner Scaler Screen Speed Element Safety Eye Wash/Shower
Spectacle Flange
6.9 Kv Switch Gear
Shredder Speed Indicator Silicon and Oxygen Analyzer 480 Volt Switch Gear
4.16 Kv Switch GearSmoke Alarm, Surface Mt. Acceler.
Smoke Detector
Snubber Solenoid (Mech. Linkage)
Sample Probe
Spacer Speed Switch (Temp. Entry)Solenoid Pilot Valve Set Press Verification Device
Square Root Extractor
Sample Rack
Signal Resistor Unit
Speed or Seismic Switch
Step Switch
Strainer Steam Unit Heater
Summer Sump Solenoid Valve
Synchroscope Meter Trap Temp Selector Switch Trip Auxiliary Unit Magnetic Tape UnitTamper Alarm Switch Terminal Box Turbidity Element Turbidity Indicating Trans Turbidity Recorder Turbidity Switch Turbidity TransmitterTemperature Controller Temperature Control Valve Time Delay Time Delay Relays Temperature Element Temperature/Moisture Element TEST THD TI TIC TIS TJR TK TM TN TNG TPA TPSA TQ TQR TQS TQT TR TRB TRC TRL TRS TS TSC TT TT/MT TUBE TV TW TY UFM USG UTD UV/OR UVD V
V/F VARM VATD VBAM VBE VBEC VBI VBIS VBR VBS VCR VD VE VIR VM VMP VPI VSC VT VTD VX VZ W
WDA WDR WDT WELL WHM WM WR WSA WSR WST WTD WUH X
XAR XAY XD XE XI XR XS XT ZONE ZSTest (MEL Diagnostics)
Thermal Detector Temperature Indicator Temperature Indicating Controller Temperature Indicating Switch Temperature Scanning Recorder Tank Timer Turn Style Turning Gear Triaxial Peak Accelerograph Testable Pipe Spool Assembly Time Totalizer Torque Recorder Torque Switch Torque TransmitterTemp./ Triax. Record./Transform.
Terminal Block Temperature Recorder Controller Translator Temperature Recording Switch Temperature Switch Temperature Scanner Temperature TransmitterTemperature/Moisture Transmitter LPRM Guide Tube Assembly Test Valve Thermal Well SMA HVAC, Special Func. Relay
Uniplex Field Module
Ultra-Sonic Generator Ultra-Sonic Transducer UV Oxidation Reactor Ultra-Violet Detector Valve Voltage/Freq. Converter Var. Meter Var. Transducer Vibration Differential AmpVibration Element Vibration/Eccentricity Indicator Vibration Indicator Vibration Indicating Switch Vibration Recorder Vibration Switch Video Cassette Recorder Viewing Device Vibration Element Vibration Instrument Rack Voltmeter Vibration Monitoring Panel Valve Pos. Indication System Variable Speed Controller Velocity Transmitter Voltage TransducerProcess Instrument Valve Vaporizer Watt Wind Direction Amplifier
Wind Direction Recorder Wind Direction Transmitter Well (For PSD System Only)
Watt Hour Meter Watt Meter Water Reprocessing Unit Wind Speed Amplifier
Wind Speed Recorder Wind Speed TransmitterWatt TransducerWater Unit Heater
Primary Containment Penetration Resid. Chlorine Analyzer Recorder Analyzer, Special Types
Explosives Detector Element, Special Types Indicator, Special TypesRecorder, Special Types Sensor, Special TypesTransmitter, Special Types
Fire Protection Zone Desig.
Tamper Switch Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.45 1.2-29.1FigureForm No. 960690Draw. No.Rev.Input (If not the Starting Point)
OutputAux Input SignalIdentification or MPL No.This block is the command switching or primary actuating function. This block can represent a switch, valve probe timer, or trip circuit. This block is normally the startingpoint of a functional sequence with an output only, but can have input and aux. input depending on the type of device. The same device may have a number of outputs, but
each functional sequence initiated shall be shown by an individual block showing the
same identification number and cross-reference. (See drawing sheet.)
Electrical power is available but the input is normallynot shown except in cases such as auxiliary power.
Battery power standby power or power from
command switches "upstream" of this block.
Location (See drawing
sheet)Signal is Present when
Condition as Described within
the Block is Met.Initiating Device Actuated
by Condition Described
within the Block.
Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.46 1.2-29.2FigureForm No. 960690Draw. No.Rev.InputOutputAux Input SignalIdentification 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
LaterNote: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 Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.47 1.2-29.3FigureForm No. 960690Draw. No.Rev.SH2-H5SH2-K7 A3This block is a permissive condition.
Where the permissive is a general condition and not identified with a single device, the outer enclosure only is shown. It only has an input and output.
If a permissive or a primary function is shown in more than one place on drawings, provide a cross-reference to the parent function. (Formally an "X" was shown in the location of
other switch handle positions, indicating that their blocks were an intricate part of the numbered switch assembly, but a different position of the switch handle. The "X" in location is inactive for new design.)
Sheet 1Sheet 2LocalEg.For F014 Parent FunctionFor F014-J5 For F014SH1-J5In this case the space is left
blank. (typical)
For F014SH1-J5(This MPL item need not be
noted if its function is obviously
apparent.)
InputOutputPermissive if general
condition as described is met.Show MPL item number of the valve or equipment served adjacent to the permissive
or primary function. (See example)
Ref. Sh. & Zone Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.48 1.2-29.4FigureForm No. 960690Draw. No.Rev.This is a seal-in with a manual reset device. The function of theseal-in is to latch in a signal and to continue that signal until manually reset. A seal-in shown without a reset device implies that the reset device is part of, and
located on the nearest valve or contactor and is automatically reset by breaking
the signal downstream of the seal-in signal. In all other cases the reset device
shall be shown in conjunction with the seal-in.This block is a final device. It can be a relay, valve, electro-mech. sw., etc.
Normally it has only inputs, but can have mech. outputs or position switch
outputs.InputResetDeviceRMSSeal-inAuxDeviceFormally this "bar"was not shown. The plain outline is inactive for new design Identificationor MPL No.Typ. of Mech. Output or Mech. Linkage Controlled Device or
Mechanism Location (See drawing
sheet)Location (See drawing
sheet or any final
device)Location (See drawing
sheet)OutputColumbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.49 1.2-29.5FigureForm No. 960690Draw. No.Rev.In this case, all signals at this point would be sealed-in.Control Sw.
in StartPositionRMSSeal-in72Examples of Typical Seal-in Blocks.
(A)Typ. Signals CRCR3-position Spring Return to NormalIn this case, only the control sw. signal would be sealed-in.Control Sw.
in StartPositionRMSSeal-in72(B)Typ. Signals CRCR3-position Spring
Return to Normal Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.50 1.2-29.6FigureForm No. 960690Draw. No.Rev.InputInputExample: description of switchgear equipment (See note)
This block is a final device used to represent motor starters, circuit breakers, etc. It hasonly input signals. The input to the right causes an opposed action to the input on the
left, such as left-open: right-close.
Note:A final device may have more than one input. Each of these inputs can initiate theblock. The block can have electrical inputs to indicating devices. Switchgear descriptions are found in ANSI spec. C37.2.
Columbia Generating StationFinal Safety Analysis Report Amendment 55 May 2001Logic Symbols for NSSSFunctional Control Diagrams 950021.51 1.2-29.7FigureForm No. 960690Draw. No.Rev.Input or Output Input or Output This block is a permissive operated by devices such as valve or pump switchgear designated in the inner block.
This condition or device effects the operation of the final device. It haselect. inputs, mech. inputs, aux. inputs (mech. or elec.), and mech. or elect. outputs. This device is normally a valve. This is also used for other input/output power sources such as air
or hydraulic.
A solenoid pilot valve for an air operated valve is an example of this type of device (see Figure 1.2-26
). When the two side blocks are the controlling blocks they have
aux. input signals.Device Title or Controlled Condition.
Aux.InputInput or Output Location (See drawing
sheet)Identification or MPL No.
Input or Output Aux InputColumbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.52 1.2-29.8FigureForm No. 960690Draw. No.Rev.This block is a primary function for ind. lights.This block is a primary function for annunciators.
XXLocation(See drawing sheet)Identificationor MPL no.
LocationLocal CR LP= =
=
= =RBWAGRedBlueWhiteAmberGreenIdentificationor MPL no.
Location(See drawing sheet)LocationLocal CR LPAnnunciator Level(s)
AL = Alarm Low AL/L = Alarm Low-low AH = Alarm High AH/H = Alarm High-high Color of LampColumbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.53 1.2-29.9FigureForm No. 960690Draw. No.Rev.StartThis line represents an elec. flow signal. This line may actuate a final device and may be used to represent
actuation of a permissive block.
This line represents an auxiliary signal source such as air or hydraulic, and is not electrical.
This line represents mechanical outputs and /or mechanical linkage.
This symbol represents the start of the primary initiating signal.This symbol represents a match circle. The letter designation on one dwg. must match the letter on the interfacing dwg.
XXXXLetter Designation Zone or Ref. Dwg.
Sheet No.
Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.54 1.2-29.10 FigureForm No. 960690Draw. No.Rev.Air Supply ExhaustTypical A.O. Valve ExampleAO F054LocalControl Sw.
in "Open" PositionRMSCRPermissive When Solenoid isEnergized Permissive When Solenoid isDe-energized Sol.PilotLocalDe-energized SolenoidPilotValveSolenoid De-energizedValve OpensExample of MPL No.
(Closed, Fails Open)
ExampleColumbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.551.2-29.11 FigureForm No. 960690Draw. No.Rev.Permissive WhenValve MO F001 Fully OpenLim. Sw.On ValveControl Sw.
In "Start" PositionRMSCRB-3B-4Permissive WhenValve MO F004 Fully OpenLim. Sw.On ValveValve MO F001 Not Fully OpenLim. Sw.On ValveValve MO F004 Not Fully OpenLim. Sw.On Valve Pump CoolingWater Temp HighTIS N002Local Pump CoolingWater Temp. HighTIS N002LocalControl Sw.
in "Stop" PositionRMSA-1Pump Discharge Flow LowFSL N003LocalControl Sw. not Permissive in "Start" PositionRMSA-1StartStopAir Circuit Breaker (52)
Signal Present When
Above Condition ExistsRGInitiating DeviceHigh Temp AlarmLocationCRAHLocationIdentificationMPL No.Permissive DeviseColumbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.56 1.2-29.12 FigureForm No. 960690Draw. No.Rev.RMSMOF001OCOCTISN002TEN001RMSMOF004OC OCDCFSLN003C001AHControl Sw. (RMS) for pump is often not shown on flow diagram but will be shown on FCD.
RMS is not Shown on FCDTypical Flow Diagram Example Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.57 1.2-29.13 FigureForm No. 960690Draw. No.Rev.AirPressureExhaustAO F033LocalEPK001LocalPermissive When Pilot Energized Permissive When Pilot De-energized F028LocalDe-energized SolenoidPilot Valve Solenoid De-energizedValve ClosesBleed-off Flow Control Valve AO F033 Functional Control Diagram PSL N013Local High Press Down Stream ofValve AO F033 PSH N014LocalValve Control Sw.
In "Open" PositionRMSCRLow Press Upstream ofValve AO F033 PSL N013Local High Press Down Stream ofValve AO F033 PSH N014Local AH/LCRE/PConverter Low Press Upstream ofValve AO F033 Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.58 1.2-29.14 FigureForm No. 960690Draw. No.Rev.Notes: 1. Aux relays and devices are not shown on FCD.Typical Flow Diagram Example F033PSHN014MainFlowPSLN013E/PRMSK001NDF028AH/LCRColumbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Logic Symbols for NSSSFunctional Control Diagrams 950021.59 1.2-29.15 FigureForm No. 960690Draw. No.Rev.Testable Check Valve AO _ _ _ _
For any Logic SeeValve Purchase Spec.For Operation of ValveSee Engr. Design Spec.Control Sw.
inPositionThis Figure is for a Typical Check Valve.Type of Sw.
"RMS"Position of Sw.VariesReason for the above change is to have one standard logic for alltestable check valve AO regardless of manufacturer.
Location:
CR, LP, Local
Command Signal Columbia Generating StationFinal Safety Analysis Report COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 prin cipal 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 d evelopment, design, constructi on, and operation of BWRs of similar types. The data, performance, charac teristics, and other info rmation presented here represent the design of the facilities at th e time of the CGS operating license review.
1.3.1.1 Nuclear Steam Supply Sy stem Design Characteristics
Table 1.3-1 summarizes the design and operating charac teristics for the nuclear steam supply systems. Parameters are related to rated pow er 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 Syst em Design Characteristics
Table 1.3-2 compares the power conversion system design characteristics.
1.3.1.3 Engineered Safety Feat ures Design Characteristics
Table 1.3-3 compares the engineered safety features design characteristics.
1.3.1.4 Containment De sign Characteristics
Table 1.3-4 compares the containmen t design characteristics.
1.3.1.5 Radioactive Waste Management Systems Design Characteristics
Table 1.3-5 compares the radioactive waste m anagement design c haracteristics.
1.3.1.6 Structural De sign Characteristics
Table 1.3-6 compares the structural design characteristics.
1.3.1.7 Electrical Power Syst ems Design Characteristics
Table 1.3-7 compares the electrical power systems design characteristics.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-3 Table 1.3-1 Comparison of Nuclear Steam Supp ly System
Design Characteristic sa CGSb BWR 5 251-764 HATCH 1c BWR 4 218-560 ZIMMERc BWR 5 218-560 Thermal a nd Hydraulic Design (see Section 4.4) Rated power (MWt) 3323 2436 2436 Design pow er (MWt) (ECCS design b asis) 3468 2550 2550 Steam flow rate (1b/hr) 14.295 x 106 10.03 x 10 6 10.477 x 106 Core coolant flow rate (1b/hr) 108.5 x 106 78.5 x 106 78.5 x 106 Feedwater flow rate (1b/h r) 14.256 x 106 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 t hermal output (KW/ft) 5.38 7.11 5.45 Maximum heat flux (Btu/hr-f t2) 428,360 428,300 354,000 Average heat flux (Btu/hr-f t2) 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 r atio (MCPR) 1.24 1.9d 1.21 Coolant ent halpy 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 pow er peaking fa ctor 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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-4 Table 1.3-1 Comparison of Nuclear Steam Supp ly System
Design Characteristic sa (Continued)
CGSb BWR 5 251-764 HATCH 1c BWR 4 218-560 ZIMMERc BWR 5 218-560 Nuclear Design (First Core)
(see Section 4.3) Water/UO2 volume ratio (cold) 2.55 2.53 2.41 Reactivity with strongest control rod out (keff) <0.99 <0.99 <0.99 Moderator void coefficient Hot, no voids (k/k - %vo id) -1.0 x 10 1.0 x 10 1.0 x 10
-3 At rated out put (k/k - %void) -1.6 x 10 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 1.3 x 10 1.3 x 10
-5 Hot, no voids (k/k - °F fuel)
-1.2 x 10 1.2 x 10 1.2 x 10
-5 At rated out put (k/k - °F fuel)
-1.3 x 10 1.3 x 10 1.3 x 10
-5 Initial average 235U enrichment wt (%)
1.91 2.23 1.90 Fuel average discharge e xposure (MW d/short ton) 10,300 19,000 15,053 Core Mecha nical 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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-5 Table 1.3-1 Comparison Of Nuclear Steam Supp ly System Design Characteristic sa (Continued)
CGSb BWR 5 251-764 HATCH 1c BWR 4 218-560 ZIMMERc BWR 5 218-560 Core Mecha nical Design (see Sections 4.2 and 7.6) (Continued)
Fuel rods (NEDE-2 0944P) 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 materia le Zircaloy-2 Zircaloy-2 Zircaloy-2 Fuel pellets
Material UO2 UO2 UO2 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 (i n.) 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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-6 Table 1.3-1 Comparison of Nuclear Steam Supp ly System Design Characteristic sa (Continued)
CGSb BWR 5 251-764 HATCH 1c BWR 4 218-560 ZIMMERc BWR 5 218-560 Core Mecha nical Design (see Sections 4.2 and 7.6) (Continued)
Reactor control system Method of v ariation of reactor power Movable c ontrol rods and variable for ced coolant flow Movable control rods and variable for ced coolant flow Movable control
rods and variable for ced coolant flow Number of movable c ontrol rods 185 137 137 Shape of movable control rods Cruciform Cruciform Cruciforn Pitch of mo vable control rods 12.0 12.0 12.0 Control material in mova ble rods B4C granules compacted in SS tubes B4C granules compacted in SS tubes B4C granules compacted in SS tubes Type of c ontrol rod drives Bottom entry locking piston Bottom entry
locking piston Bottom entry
locking piston Type of tem porary reactivity control for initial co re Burnable poison; gadoliniaur ania fuel rods Burnable poison; gadoliniaur ania fuel rods Burnable poison; gadoliniaur ania fuel rods In-core neut ron instrumen tation Number of in-core neutron detectors (fixed) 172 124 124 Number of in-core detect or assemblies 43 31 31 Number of d etectors per a ssembly 4 4 4 Number of flux mappi ng neutron detectors 5 4 4 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-7 Table 1.3-1 Comparison of Nuclear Steam Supp ly System
Design Characteristic sa (Continued)
CGSb BWR 5 251-764 HATCH 1c BWR 4 218-560 ZIMMERc BWR 5 218-560 Core Mecha nical Design (see Sections 4.2 and 7.6) (Continued)
In-core neut ron instrumen tation (Continued)
Range (and number) of de tectors Source range monitor Source to 0.001% po wer (4)f Source to
0.001% po wer (4)f Source to
0.001% po wer (4)f Intermediate range m onitor 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 m onitor 2.5% to 1 25% power (6)f 2.5% to 1 25%
power (6)f 2.5% to 1 25%
power (6)f Number and type of in-co re neutron sources 7 Sb-Be 5 Sb-Be 5 Sb-Be Reactor Ves sel Design (see Section 5.3) Material Carbon steel stainless clad Carbon steel
stainless clad Carbon steel
stainless clad Design pres sure (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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-8 Table 1.3-1 Comparison Of Nuclear Steam Supp ly System Design Characteristic sa (Continued)
CGSb BWR 5 251-764 HATCH 1c BWR 4 218-560 ZIMMERc BWR 5 218-560 Reactor Coo lant Recircul ation Design (see Sections 5.1, 5.2, and 5.4) Number of recirculation loops 2 2 2 Design pres sure: Inlet leg (psig) 1250 1148 1250 Outlet leg (psig) 1650;g 1550h 1274 1675;g 1575h Design temperature (°F) 575 562 575 Pipe diamet er (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 r eactor 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-9 Table 1.3-2 Comparison of Power Conversion System Design Characteristics CGS BWR 5 251-764 HATCH Ia BWR 4 218-560 ZIMMERa BWR 5 218-560 Turbine Generator (see Sections 10.2 and 10.4) Rated power (MWt) 3468b 2550 2550 Rated power (MWe) (gros s) 1205b 813 883 Generator S peed (rpm) 1800 1800 1800 Rated steam flow (1b/hr) 15.018 x 106b 10.48 x 10 6 11.0 x 106 Inlet pressure (psia) 955 950 950 Steam Bypass System (see Section 10.4.4) Capacity (% design steam flow) 25 25 25 Main Conde nser (see Section 10.4.1) Heat removal capacity (B tu/hr) 7702 x 106 5720 x 10 6 7053 x 106 Circulating Water System (see Section 10.4.5) Number of p umps 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 106 10.096 x 10 6 10.971 x 106 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-10 Table 1.3-3 Comparison of Engi neered 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 spr ay system Number of loops 1 1a 1 Flow rate (gpm) 1550 at 1130 psid 4250 1330 at 1110 psid 6350 at 200 psid 4725 at 200 psid Automatic depressur ization sys tem Number of relief valves 7 7 7 Low pressure coolant injectio nb 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 106 32 x 106 30.8 x 106 Number of heat exchangers 2 2 2 Primary containment cooling mode:
Flow rate (gpm) 7450 e 30,800 5050 e
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-11 Table 1.3-3 Comparison of Engi neered 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 3f 4 4 Reactor Core Isol ation Cooling System (see Section 5.4.6) Flow rate (gpm) 600 at 1150 psid 400 at 1120 psid 400 at 1120 psid Fuel Pool Cool ing and Cleanup Sy stem (see Section 9.1.3) Capacity (Btu/hr) 8.0 x 106 5.7 x 106 6.6 x 10 6 a High-pressure coolant injection system utilized.
b A mode of RHR system.
c Capacity during reac tor flooding mode with mo re 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-12 Table 1.3-4 Comparison of Containme nt Design Characteristics CGS BWR 5 251-764 HATCH 1 BWR 4 218-560 ZIMMER BWR 5 218-560 Primary Con tainmenta (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 pr e-stressed with
steel liner Drywell Frustum of cone upper portion Light bulb/s teel vessel Frustum of cone
upper portion Pressure-su ppression ch amber Cylindrical lower portion
with eliptical
bottom Torus/steel vessel Cylindrical
lower portion Pressure-su ppression ch amber internal
design pressure (psig) 45 56 45 Pressure-su ppression ch amber extern al design pressure (psi) 2 2 2 Drywell internal design p ressure (psig) 45 56 45 Drywell external design p ressure (psi) 2 2 2 Drywell free volume (f t3) 200,540b 146,240 180,000 Pressure-suppression chamber free volume (ft3) 144,184 max 110,950 93,000 Pressure-su ppression p ool water volume
(ft3) 112,197 minc 87,300 102,000 Submergence of downco mer vent pipe below press ure pool surf ace (ft) 12 max.
11.67 min.
3.67 10 Design temperature of dr ywell (°F) 340 281 340 Design temperature of pre ssure-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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-13 Table 1.3-4 Comparison of Containment Des ign Characteristics (Continued)
CGS BWR 5 251-764 HATCH 1 BWR 4 218-560 ZIMMER BWR 5 218-560 Primary Con tainmenta (see Sections 3.8.2 and 6.2.2) (Continued)
Calculated maximum pres sure after bl owdown to dwell (no pre-surge) (
psig) 34.7 46.5 40.4 Pressure-su ppression ch amber (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 desi gn pressure (in. H 2O) 0.25 0.25 0.25 Design inleakage rate (% free volume/day at 0.25 in. H 2O) 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 with in the reactor pedestal (10,065 ft
- 3) or the 12 ft of water below the downcomer vent pipe exits (15,000 ft 3).
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 Ra dwaste (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 trea tment Low temperature
charcoal Recombiner
ambient charcoal Chilled charcoal Number of b eds 8 12 5 Design cond enser in-leak age (cfm) 30 40 12.5 Release point
- height ab ove ground (ft) 230 394 172 Liquid Rad waste (see Section 11.2) Treatment of
- 1. Floor dr ainsa F, D, and R F, D, and R F, E, and R 2. Equipm ent drainsa F, D, and R F, D, and R F, D, and R
- 3. Chemical drain sa Neutralized, E, D, and R F, discharg ed E, solid to
radwaste E, D, concentrates to
solid radwaste
distillate R
- 4. Detergent drain sa Chemical addition, F, E, and sent to
circulating
water dischargeb Diluted and sent to circulating
water discharge Reverse osmosis discharge
- 5. Expect ed 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 (ho rizontal g) 0.250 0.15 0.20 Wind Design (see Sect ion 3.3) Maximum sustained (mph) 100 105 90 Tornados Translational (mph) 60 60 60 Tangential (mph) 300 300 300
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-16 Table 1.3-7 Comparison of Electrical Sy stems Design Characteristics CGSa 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/ba ckup auxiliary transformers 2 2 2 Standby ac power supp ly Number of die sel generators 3b 3c 3 Number of 4160-V shu tdown (Class 1E) buses 3b 3 3 Number of 480-V shutdown (Class 1E) buses 5b 2 (600 V) 5 Power Supply (d c) (see Section 8.3.2) Number of 24-V batteries 4 2 (48 V) Number of 125-V batteries 6d 3 3 Number of 250-V batteries 1 2 1 Number of 24-V bu ses 2 2 (24/48 V) Number of 125-V buses 6d 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.
COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000LDCN-99-000 1.3-17Table 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-18Table 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 COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000LDCN-99-000 1.3-19Table 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.
(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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-20Table 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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-21Table 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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.3-22Table 1.3-8 Significant Design Changes from PSAR to FSAR (Continu ed) Item Change Reason for Change FSAR Portion in Which Change is Discussed Offgas holdup line Radiography of circumferent ial welds was not done. A partial section of the line was buried before radio graphy was done. Welds were
magnetic part icle tested and line was hydro-tested at 1200 psig and then helium pressure decay leak tested wi th a sensitivity of 10-2 cm3/sec. - Wet solid ra dwastes Packaged in 50 ft3 containers rather than 50-gal drums.
Reduce handl ing time and operator exposure.
11.4.2.10 Turbine bypass valve
system Four bypass valves are used rather than th ree. Solution to operating proble ms with bypass valves on Co oper Nucle ar station. 10.4.4 Main steam isolation valve leakage control system Added to plant.
NRC requirement. 6.7 Main steam li ne from outermost iso lation valve to turbi ne stop valve Piping has been upgraded fr om Code Group D to Code Group B.
NRC require ment. 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 COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000LDCN-99-000 1.3-23Table 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 COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000LDCN-99-000 1.3-24Table 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 hooks tick and 230-kV OCB at plant switchy ard. 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 av ailability.
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 COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORTDecember 2011LDCN-11-008 1.3-25Table 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.
COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT 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 providi ng 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 th e 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 ge nerating capacity, of which more than 4,800,000 kW is nuclear.
Burns & Roe, Inc., has been continuously engaged in construction of e ngineering activities since 1935.
1.4.3 NUCLEAR STEAM SYSTEM SUPPLIER - GENERAL ELECTRIC COMPANY
General Electric designed, f abricated, and delivered the direct-cycle boiling water nuclear steam supply system (NSSS) for Columbia Generating Station (C GS). General Electric also fabricated the first core of nuc lear fuel and provided technica l direction of installation and startup of this equipment.
General Electric has engaged in the developm ent, design, construction, and operation of boiling water reactors (BWRs) since 1955. Table 1.4-1 lists GE reactors completed, under construction, or ordere d (several later canceled). Thus, GE has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the installation and star tup of reactors.
General Electric continues to provide technical support for th e operation of CGS as requested by Energy Northwest. This includes providing support for the CGS Megawatt Improvement Program (see Section 1.1).
COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 1.4-2 1.4.4 TURBINE GENERATOR SUPPLIER
- WESTINGHOUSE ELECTRIC CORP.
Westinghouse Electric Corporati on designed, fabricated, deliver ed, and installed the turbine generator for CGS. They also provided technica l assistance for the start up of this equipment.
Westinghouse Electric Corporati on has a long history in the app lication of turbine generators in nuclear power stations going back to the inception of commercial electrical power production using nucle ar facilities. Westinghouse furnis hed 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 turbin e generators produced over 300 billion kW hr of electricity through May 1976, when 25 nucle ar turbine generators totaling over 16,500 MW were in serv ice. 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 provid e technical and maintenance suppor t 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, pipelin es, refineries, and aircraft modification. Next, Bechtel pioneered in the nuclear power field, constructing the first reacto r to produce useful electricity in 1949, and building Dresden I, the first commercial nucle ar power station. Today, Bechtel is recognized as one of the world's leading engineering and construction firms.
1.4.6 MAJOR CONTRACTORS 1.4.6.1 Fischbach/Lord
Fischbach/Lord is responsible for the major electri cal installation at CGS, consisting of raceways, conduit, cable, termi nators, and electrical equipment. They were formed as a joint venture, solely for th is project, in 1974.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December2007 LDCN-07-011 1.4-3 1.4.6.2 Pittsburgh-Des Moines Steel Company
Pittsburgh-Des Moines Steel Co mpany is responsible for e ngineering, fabrication, and installation of materials in the Primary Containment Vessel.
1.4.6.3 Wright-Schuchart-Harbor/Boecon (Boeing Cons truction)/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, Bechte l served as the Construction Manager. During the operating phase Bechtel, as the Site Support Services c ontractor, is providi ng field engineering and installation support for plant mod ifications. Also, as Technical Services contractor, they are providing engineering support under Energy Northwest dire ction and under the Energy Northwest quality assurance program as request ed by Energy Northwest. Under these contracts Bechtel is providing support to the Megawatt Improvement Program (see Section 1.1).
1.4.6.5 AREVA NP
The initial fuel core wa s fabricated by GE. Reload fuel is being provided by AREVA NP.
Their contract provides for the supply of uraniu m concentrates and fuel fabrication services.
Other fuel in the core was provided by Westinghouse (ABB/Combustion Engineering).
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. Westinghous e 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 c onsulting engineer for Energy Northwest's Columbia Generating Station. This firm was also a consulting engineer for WNP-1. Having extensive experience in preparing engi neering feasibility and financing studies and reports necessary for the success of utility and civic im provement projects, the firm is well qualified for employment as a consulting engineer and was chosen as a result of its experience.
COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT 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 feas ibility of the project. These duties will include assisting in pr eparation of a Bond Resolution, preparation of an engineering report, schedules for investme nt of funds, schedules for de bt service payments, and other engineering services necessary to fa cilitate the financing of the project.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.4-5 Table 1.4-1 Commercial Nuclear Reactors C ompleted, Under Construction, or in Design by General Electric
Station Utility Rating (MWe) Year of Order Year of Startup Dresden 1 a Commonweal th 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 Moh awk 610 1963 1969 Dresden 2 Commonweal th Edison 794 1965 1970 Pilgrim 1 Boston Ed ison 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 Commonweal th Edison 794 1966 1971 Monticello Northern States 536 1966 1971 Quad Cities 1 Commonweal th Edison 789 1966 1972 Browns Ferry 1 TVA 1065 1966 1974 Browns Ferry 2 TVA 1065 1966 1975 Quad Cities 2 Commonweal th 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 Auth ority 821 1966 1975 Bailly b NIPSCo 660 1966 ---- Shoreham b LILCo 819 1967 1985 Cooper Nebr aska 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 ----
COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 LDCN-99-000 1.4-6 Table 1.4-1 Commercial Nuclear Reactors C ompleted, Under Construction, or in Design by Gen eral 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 Commonweal th Edison 1078 1970 1983 LaSalle County 2 Commonweal th 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 Ge nerating Station Energy Northwest 1103 1971 1984 Nine Mile Point 2 Niagara Moh awk 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 E lectric 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 COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT 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 addresse d as unresolved safety issues (USIs), generic safety issues (GSIs), and TMI Task Action Plans.
Human Factors consider ations are included as part of the TMI Task Action Plans. Cher nobyl is not addressed belo w or on the CGS docket as NUREG-1251 lead to the c onclusion 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, pro cedures, or training at CGS.
COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 LDCN-99-000 1.5-2 1.5.1.1 Unresolved Safety Issues
1.5.1.1.1 Unresolved Saf ety Issues Introduction
Unresolved safety issues are issues identified by the NRC that affect a number of plants, question the adequacy of existi ng 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 Ener gy Reorganization Act required that the NRC develop a plan providing for speci fication 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 Gene ric Letter 89-21, "Request for Information Concerning Status of Implementa tion of Unresolved Safety Issu e (USI) Requirements." This generic letter requested that licensees and c onstruction permit holders review and report on the status of the implementation of USIs for wh ich final technical reso lution 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 Sy stems (A-9, A-44, and A-47, respectively) as not being implem ented. (Subsequently, these ha ve been resolved as discussed below.)
1.5.1.1.2 Implementation of Spec ific 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 Sect ions 6.2.1.8 and 3.9.3.1, respect ively.
A-9 Anticipated Trans ients 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 COLUMBIA GENERATING STATION Amendment 56 FINAL SAFETY ANALYSIS REPORT 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 alte rnate rod injection system was in compliance with the ATWS rule. The reference also stated that the RPT system required two modifications to be in compli ance with the rule. Reference 1.5-4 documents the implementation of the changes require d to resolve th ese two issues.
In Reference 1.5-5, Energy Northwest informed the NRC that confirmation of the environmental qualifications of ATWS equipmen t 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 ne w ATWS analyses. Technical Specification Amendment 93 was issued on August 9, 1991 which addresse d modifications to the ATWS-RPT system. With th is amendment, all activities re quired 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 Generi c 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 rais ed in the subsequent five supplements to NUREG-0892 and as Energy Northwest was not aware of a concern of the NRC's 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 ha ve been required to submit a fracture analysis to justify continued operation. This wa s not the case for CGS.
A-17 Systems Interactions
Generic Letter 89-18 issued Se ptember 6, 1989 transmitted NRC final resolution of this USI.
No formal reply was required. Energy Nort hwest incorporated information contained and referenced in this Generic Le tter into the CGS IPE program
, the results of which were COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT 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 me morandum, 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 Commission's regulations and applicable Regulatory Guides. Ge neric 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 establishe d closure of A-31 for CGS.
NUREG-0892 Supplement 4, Sectio n 9.1.5, states that the gu idelines of NUREG-0612 have been satisfied for CGS. Generic Letter 89
-21 states that NUREG-0612 resolves A-36.
A-39 Determination of Safe ty Relief Valve Pool Dynamic Load and Temperature Limits
Section 6.2.1.8 of NUREG-089 2 Supplements 1 and 4, provid es 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 applic ants only. CGS is not one of the plants identified in Generic Letter 89-21 that need ed to be reviewed to the new criteria.
A-42 Pipe Cracks in Boiling Water Reactors NUREG-0892 states in Secti on 5.2.3.1 that CG S 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 COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT 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 Emer gency Sump Performance
Generic Letter 89-21 states that resolution of A-43 only applie s 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, Enclos ure 1, established Generic Letter 87-03 as applicable to CGS rath er than Generic Letter 87-02.
As such, Energy Northwest considers this USI closed for CGS. Also, NUREG-0892, Suppl ement 5 in Appendix C states that A-46 only applies to plants t hat 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 fo r in CGS. Closure of this is sue was provided by Reference 1.5-9.
A-48 Hydrogen Control Measur es and Effects of Hydrogen Burn on Safety Equipment
As stated in Generic Letter 89-21, A-48 is closed and implemen ted 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 NR C. Regarding Station Blackout (A-44), 10 CFR 50.63(c
)(4) provides for a 2 year impl ementation schedule for closure of identified modifications.
COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 LDCN-99-000 1.5-6 1.5.1.2 Generic Safety Issues
1.5.1.2.1 Generic Safet y 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 Northwest's 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 GS Is listed in Generic Letter 90-04 and other GSIs that have be en resolved for CGS subseque nt to the issuance of the Generic Letter. The followi ng is a summary of informa tion 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 Ge neric Letter 90-04 response)
Closed as documented in Reference 1.5-19 49/Interlocks and LCOs for Class 1E Tie Breakers - Generic Le tter 91-11 (added subsequent to Ge neric 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 COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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, Su pplement 1, issued October 7, 1992, did no t 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 Techni cal Specifications were based upon Revision 3 to the BWR STS
issued in 1980, this c oncern is resolved for CGS. In particular, for the five issues
mentioned for GSI A-13 resolution in Generic Letter 90-04: 1. The arbitrary c apacity 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. Th ey are significantly expanded from that included in earlier programs.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 Ge neric 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 discu ssed in Reference 1.5-13) B-63/Installation of Low Pressure Systems Connected to the RCPB Closed as discusse d 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 include d 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, "Antic ipated Transients Without Scram (ATWS) Design Modifications," dated June 22, 1990.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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 J une 29, 1990.
1.5-7 Letter, R. B. Samwort h (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 HP CS 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 Issu e 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, "Ser vice Water System Problem s 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 J une 28, 1990.
1.5-14 Letter, G. W. Kni ghton (NRC) to G. C. Sorensen (SS), "Emergency Response Capability - Conformance to Regulator y 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 Suppl y Problems Affecting Safe ty-Related Equipment,"
dated July 28, 1989.
COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT 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), "Respons e 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), "Respons e 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 Powe r 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. 69 161)," dated January 8, 1992.
1.5-36 Letter, GO2-92-086, G.
C. Sorensen to NRC, "Add itional Response to Generic Letter 88-01 Safety Evaluation Report (T AC Nos. M80358 and M69161)," dated April 10, 1992.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCE Table 1.6-1 is a list of General Electric topical repor ts and other reports and documents which are incorporated in whole or in part by referen ce. These documents were filed with the NRC.
COLUMBIA GENERATING STATION Amendment53 FINAL SAFETY ANALYSIS REPORT November1998
Table 1.6-1 Topical Reports Report Title FSAR Section 1.6-3 General Electric Company APED-4824 Maximum Two-Phas e Vessel Blowdown from Pipes (April 1965) 6.2 APED-5458 Effectiveness of Core Standby Cooling Systems for General El ectric Boiling Water Reactors (March 1968) 5.4 APED-5460 Design and Performa nce 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 Considerati ons in Design of Large Boiling Water Reactors (June 1968) 4.1 APED-5652 Stability and Dyna mic Performance of the General Electric Boili ng 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 Bo iling Water Reactors (November 1968; revised April 1969) 7.6 APED-5750 Design and Performa nce 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 COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section LDCN-08-035 1.6-4 GEAP-10546 Theory Report for Creep-Plast Computer Program (January 1972) 4.1 GEAP-13197 Emergency C ooling in BWRs Under Simulated Loss-of-Coolant (BWR PLECMP)
Final Report (June 1971) 6.2 GE-NE-778-028-0790 GE Duralif e 215 Control Rod Safety Evaluation, Revisi on 2 (July 1992) 4.2 GE-NE-187-24-0992 Washington Pub lic Power Supply System Nuclear Project 2, SRV Setpoint Tolerance
and Out-of-Service Analysis, Revision 2
(July 1993) 6.3 NEDC-31984-P Generic Evaluati ons of Genera l Electric Boiling Water Reactor Power Uprate -
(July 1991) 5.4, 15.8 NEDC-32115-P Washington Pub lic 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 Gene ral Electric Met hodology for Reactor Pressure Vessel Fast Neutron Flux
Evaluations (J anuary 2006) 4.3.2.8, 4.3.4 NEDE-10169 Safe-System Analysis for Standby Core Cooling Equipment (September 1970) 3A COLUMBIA GENERATING STATION Amendment59 FINAL SAFETY ANALYSIS REPORT December2007
Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section LDCN-07-011 1.6-5 NEDE-10313-P PDA - Pipe Dyna mic 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 - Pressu re Suppression Test Program (May 1976) 3A NEDE-20943-P Urania-Gadolinia Nuclear Fuel Physical and Irradiation Characteristics and Material Properties (January 1977) 4.2 NEDE-20944-P BWR/4 a nd 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 Cha nnel 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 No zzle/Sparger Final Report (March 1978) 5.2, 5.3 NEDE-23604 Brunswick Unit 1 Reacor Internals Vibration and Temperature Measur ements (June 1977) 5.3 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section LDCN-13-013 1.6-6 NEDE-23749-P Analytical Mode l for Computing Transient Pressure and Forces in the S/RVDL (February 1978) 3.9 NEDE-23806-P MK II Main Ve nt Lateral Loads Summary Report (October 1978) 3A NEDE-24010-P Technical Base s 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 Sta ndard 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 Plan ts (November 1977) 3.9 NEDE-24106-P Dynamic Latera l Loads on a Main Vent Downcomer - Mark II Containment (March 1978) 3A NEDE-24222 Assessment of Boiling Water Reactor Mitigation of Anticipat ed 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 Condens ation Oscillation Load Definition Report (November 1980) 3A NEDE-24302-P Generic Chugging Load Definition Report (April 1981) 3A COLUMBIA GENERATING STATION Amendment59 FINAL SAFETY ANALYSIS REPORT December2007 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section LDCN-06-000 1.6-7 NEDE-24695 RVF0R04 User's 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 Condensati on Oscillation Test Program Final Test Report (May 1980) 3A NEDE-24822-P Mark II Impr oved Chugging Methodology (May 1980) 3A NEDE-24834 Hanford 2 Crimpe d Control Rod Drive Line (June 1980) 3.6 NEDE-24988-P Analysis of Gene ric BWR Safety/Relief Valve Operability Test Results (October 1981) 5.2, 5.4, Table F.4-1 NEDE-25100-P CAORSO SRV Disc harge Tests Phase I Test Report (May 1979) 3A NEDE-25118 CAORSO SRV Disc harge Tests Phase II ATR Report (August 1979) 3A NEDE-31096-P Licensing Topi cal Report, Anticipated Transient Without Scram (February 1987) 4.6, 7.4, 9.3 NEDM-10320 The General Elec tric Pressure Suppression Containment Analytical Model (March 1971) 3A, 6.2 NEDO-10029 An Analytical St udy on Brittle Fracture of GE BWR Vessel Subject to the Design Basis
Accident (July 1969) 1.8 NEDO-10320 The General Elec tric Pressure Suppression Containment Analytical Model (April 1971) 3A 6.2 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 Antic ipated Transients Without Scram (March 1971) 15.8 NEDO-10466-A Power Generati on Control Complex Design Criteria and Safety Evaluation
(September 1977) 8.3, 9.5, F.2, F.3, F.7 NEDO-10527 Rod Drop Acci dent 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 Genera l Justification for Classification of Effluent Treatment System Equipment as Group D (February 1973) 11.3 NEDO-10751 Experimental a nd Operational Confirmation of Off-Gas System Design Parameters (January 1973) (Proprietary) 11.3 NEDO-10802 Analytical Me thods of Plant Transient Evaluations for Genera l Electric Boiling Water Reactor (February 1973) 15.2 NEDO-10899 Chloride C ontrol in BWR Coolants (June 1973) 1.8, 5.2 NEDO-10905 HPCS Power Supply 1.8, 8.3 COLUMBIA GENERATING STATION Amendment59 FINAL SAFETY ANALYSIS REPORT December2007 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section LDCN-05-009 1.6-9 NEDO-10951 Releases from BWR Radwaste Management Systems (July 1973) 11.2 NEDO-10958-A General Electr ic BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application (January 1977) 6A NEDO-20533 The General Electric Mark III Pressure Suppression Containment System Analytical
Model (June 1974) 3A, 6.2 NEDO-20566-P-A Analytical Model for Loss-of-Coolant Analysis in Accord ance with 10 CFR 50, Appendix K (Proprieta ry) (January 1976) 3.9, 4.2 NEDO-20626 Studies of BWR Designs for Mitigation of Anticipated Transients without Scrams
(October 1974) 6.2, 9.3 NEDO-20761 Millstone Nu clear Power Station, Refueling/Maintenanc e Outage (Fall 1974) 12.2 NEDO-21061 Mark II Contai nment Dynamics Forcing Functions Information Report
(September 1976, June 1978, November 1981) 3A, 6.2 NEDO-21142 Realistic Accide nt Analysis for General Electric Boiling Water Reactor - The RELAC
Code and User's Gu ide (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 COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT 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 Capab ility 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 Plan ts (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 Anal ysis of NRC Problem (August 1979) 3.9 NEDO-24226 General Electric Company, Control Blade Lifetime With Potential B 4C Loss, with Supplement 1 (December 1979) 4.2 NEDO-24288 Mark II Contai nment Program - Generic Condensation Oscillation Load Definition Report (February 1981) 3A NEDO-24548 Technical Descri ption Annulus Pressurization Load Adequacy Evaluation (January 1979) 6.2 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section LDCN-12-021 1.6-11 NEDO-24708-A Additional Info rmation Required for NRC Staff Generic Repor t on Boiling Water Reactors (June 1980) 7.4, B, I, Table F.4-1 NEDC-24154-P-A Qualif ication of the One-Dimensional Core Transient Model for Boiling Water Reactors, Volumes 1, 2, 3 and 4 (February 2000) 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 Met hodology and Uncertain ties for Safety Limit MCPR Evaluations (August 1999) 4.4 NEDC-32694P-A Power Distributi on Uncertainties for Safety Limit MCPR Evaluations (August 1999) 4.4 NEDC-32851-P-A GEXL14 Correlation for GE14 Fuel (April 2011) 4.4 NEDC-32868P GE14 Complianc e With Amendment 22 of NEDE-24011-P-A (GESTAR II) (January 2012) 4.1, 4.2, 4.3, 15.4 NEDC-32950P Compilation of Improvements to GENE's SAFER ECCSLOCA Evaluation Model (July 2007) 6.3 NEDC-33419P GEXL97 Corre lation Applicable to ATRIUM-10 Fuel (June 2008) 4.4 NEDE-23785-1-PA The GESTR-LO CA and SAFER Models for the Evaluation of the Loss-of-Coolant
Accident. Volumes 1, 2, and 3
(October 1984) 6.3 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section LDCN-13-013 1.6-12 NEDE-23785P-A The GESTR-LO CA 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 Sta ndard 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 Exxon Nuclear Company / Advanced Nuclear Fuels Corp. / Siemens Power Corporation / Framatome ANP / Areva NP Inc. ANF-524 (P) (A) Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors, Revision 2, and Supplements 1 and 2 (November 1990) 4.4 ANF-913(P)(A) CONTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4 (August 1990) 15.1, 15.2 ANF-1358(P)(A) The Loss of Feed water Heating Transient in Boiling Water Reactors, Revision 3
(September 2005) 15.1 ANF-89-98 (P)(A) Generic Mechan ical Design Criteria for BWR Fuel Designs, Revision 1 and Supplement 1 (May 1995) 3.9, 4.1, 4.2, 4.3, 4.4 EMF-CC-074 (P)(A) BWR Stability Analysis Assessment of STAIF with Input from MICROBURN-B2,
Volume 4, Revision 0, (August 2000) 4.1, 4.3 COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section 1.6-13 EMF-93-177 (P)(A) Mechanical De sign for BWR Fuel Channels, Revision 1 (August 2005) 3.9 EMF-2158(P)(A) Siemens Power Corporation Methodology for Boiling Water Reactors; Evaluation and
Validation of CASMO-4/MICROBURN-B2,
Revision 0 (October 1999) 4.4, 15.1, 15.4 EMF-2209(P)(A) SPCB Critical Power Correlation, Revision 2 (September 2003) 4.4 EMF-2245(P)(A) Applications of Siemens Power Corporation Critical Power Correlati ons to Co-resident Fuel, Revision 0 (August 2000) 4.4 XN-NF-80-19 (P)(A) Exxon Nucl ear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Su mmary Description, Volume 3, Revision 2 (January 1987)
Application of the ENC Methodology to BWR
Reloads, Volume 4, Revision 1 (June 1986) 15.4
15.4 XN-NF-81-58 (P)(A) RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation M odel, Revision 2 and Supplements 1 and 2 (March 1984) 6.3 XN-NF-82-07 (P)(A) Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, Revision 1
(November 1982) 4.2, 6.3 Asea Brown Boveri (ABB) / CE Nuclear Power / Westinghouse Electric Company CENPD-287-P-A Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors
(July 1996) 4.4 COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section 1.6-14 CENPD-300-P-A Reference Safe ty Report for Boiling Water Reactor Reload Fuel (July 1996) 4.1, 4.2, 4.3, 4.4 CENPD-392-P-A 10 x 10 SVEA Fuel Critical Power Experiments and CP R Correlations:
SVEA-96, Revision 0 (September 2000) 4.4 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
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 Wa ll, Supplement No. 1 3.8, 6.2 -- Plant Design Assessment Report for SRV and LOCA Loads 3A WPPSS-74-2-R3 Burns & Roe, In c., 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 Prog ram Plan Interval - 2 5.2.4, 6.6 COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 Table 1.6-1 Topical Reports (Continued)
Report Title FSAR Section 1.6-15 Preservice Inspection
Program Plan Preservice Inspection Program Plan 5.2.4, 6.6 CGS-FTS-0168 Columbia Gene rating Station Alternative Source Term (report consolidated from letters
GO2-04-170 dated Se ptember 30, 2004,
GO2-06-116 dated Se ptember 11, 2006,
GO2-05-054 dated March 16, 2005,
GO2-05-160 dated Se ptember 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 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 tr ansfer, 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
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 ch ange authorization
ECCS emergency core cooling system
ECN engineering change notice
EDR equipment drain (radioactive) processing COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 Powe r Research Institute
EPZ emergency planning zone EQ Environmental Qualificati on, 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 Deviati on Disposition Request
FDR floor drain (radioac tive) processing system
FLECHT full-length emergenc y cooling heat transfer COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-09-035 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 HELB high energy line break
HEPA high-efficiency particulate air/absolute
HID high-intensity discharg e (lighting--vapor lamp)
HPCS high-pressure core spray
H&V heating and ventilating
HVAC heating, ventilati ng, and air conditioning
HX heat exchanger IASCC Irradiation Assist ed 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
IGSCC intergranular stress corrosion cracking COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-09-035 1.7-6 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
MG motor-generator COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-09-035 1.7-7 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 c ontrol system
msl mean sea level
MSLC main steam isolat ion 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 tr ansition 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
NPDES National Pollutant Discharge Elimination System COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-09-035 1.7-8 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 Sa fety and Health Act
OS&Y outside screw and yoke
PA Public Address (System)
PABX Private Automatic Branch Exchange
PATP Power Ascension Test Program
PCIOMR preconditioning cladding interi m operating management recommendation
PCRVICS primary containment and reactor vessel isolation control system
PCS process computer system COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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
RHR residual heat removal RMC reactor manual control
RMS remote manual switches
ROA reactor building outside air COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-09-035 1.7-10 RPIS rod position information system
RPT recirculation pump trip
RRC reactor recirculation system RRS required response spectra
RSCS rod sequence control system
RSO reactor system outline
RTNDT Reference Temperature for Nil-Ductility Transition RWCU reactor water cleanup
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 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-09-035 1.7-11 SCC/IASCC SCC/irradiation assisted stress corrosion cracking
SEF single equipment failure
SER Safety Evaluation Report
SF single failure (NSOA)
SGT standby gas treatment
SGTS standby gas treatment system
SLMCPR minimum critical pow er 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 stainless steel
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-09-035 1.7-12 SSC structures, systems, and components
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 (tur bine building service water)
TWG Test Working Group
UBC Uniform Building Code COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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
COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT 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 informa tion on Energy Northwest's conformance assessment of CGS to Regulatory Guides, Division 1, Powe r Reactor Guides and revisions thereof as noted.
Since the scope of equipment re sponsibility is project unique and the time of equipment design, procurement, manufacture, installation, and op eration varies with the supplier, a unique assessment for the nuclear stea m supply system (NSSS) scope of supply and balance of plant (BOP) scope of supply is n ecessary and is presented.
1.8.2 NUCLEAR STEAM SUPPLY SYSTEM SCOPE OF SUPPLY EVALUATION
The following paragraphs define the nomenclatur e and the manner in which the NSSS scope of supply assessment is to be interpreted.
Regulatory Guides - Inco rporated in the Design
This section serves to identify specific safe ty 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 af ter the construction pe rmit issuance. All of these are specifically noted as "Incorporated in the Design."
Regulatory Guides - Assessed Capability in the Design
For those other regulatory guide s which have been issued be fore, during, or after the construction permit issuance, Energy Northwest (through his agents and/or suppliers) has performed an assessment evaluation to de termine 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 requi rements as described in the FSAR.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 wi th them is not requi red. 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 continuan ce 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 represen ts the NSSS scope of supply regulatory guide assessment.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.8-3 Regulatory Guide 1.1, Re vision 0, November 1970 Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal
Pumps.
Regulatory Guide Intent:
This guide prohibits design re liance on pressure and/or temperature transients expected during a loss-of-coolant accid ent (LOCA) for ensuring adequate net positive suction head (NPSH). The requiremen ts of this regulatory guide are applicable to the high-pressure core spray (HPCS), low-pressu re 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, desi gn, 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 cons ervatively 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.
COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 1.8-4 Regulatory Guide 1.2, Re vision 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 significan t 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 anneali ng, could be applied to ensure adequate recovery of the fracture toughness properties of the vessel materi al. This regulatory guide requires that available engineering solutions be outlin ed 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, desi gn, and/or equipment used in CGS is in compliance with the in tent 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 Ve ssel Subject to the Design Basis Accident"). It has been determined, based on met hods of fracture mechanics, th at 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 effe ct of blowdown and the LPCI system reflooding,
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 resi dual stresses,
- c. The radiation effect on mate rial 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 fi eld following DBA.
This analysis incorporated very conservative assumptions in all areas (partic ularly in the areas of heat transfer, stre ss analysis effects of radia tion on material toughness, and crack tip stress intensity).
Therefore, the results reporte d in NEDO-10029 provide an upper bound limit on brittle fractu re failure mode studies. Because of the upper bound approach, it is concluded that catastrophic failu re of the pressure vessel due to the DBA is shown to be impossible from a fracture m echanics 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 estab lishing 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 paragra ph NB2331 of the ASME Secti on III Code. This predicted value is not exceeded; therefore desi gn for annealing is not required.
Specific Evaluation
Reference:
See Section 5.3.1.5. Similar Application
Reference:
Similar application was used fo r Browns Ferry 1, 2, and 3.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-04-050,05-055 1.8-6 Regulatory Guide 1.6, Revision 0, March 1971
Independence Between Redunda nt 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 Genera l 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 an alysis, design, and/or equipm ent 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 automa tically 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 c ontrolled in the open position to prevent paralleling of standby sources.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 batte ry charger and one battery.
The HPCS standby power source and distributi on system is independe nt from the other two standby power sources and associated distributi on system in the plant.
Specific Evaluation Reference
See Section 8.3.1.2. Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 ensu ring sufficient onsite pow er capability and for determining load requirements of di esel generator set power sources.
Application Assessment:
Incorporated in design.
Compliance or Alternate Approach Statement:
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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 coincident al 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-9 Regulatory Guide 1.9, Position 4 Conformance The HPCS diesel generator un it 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 tw o to three times as large as that required to carry the continuous rated load would be necessary
. The specific dies el 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-s tart 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 dies el-engine generator unit and all of the major system auxiliaries.
GE Licensing Topical Report, HPCS Po wer Supply, NEDO-1090 5, describes the theoretical analytical aspects of the un ique application including prototype and reliability test considerations.
The design of the HPCS diesel generator conf orms with the applicable sections of IEEE criteria for Class 1E "Electrical Systems for Nuclear Power Ge neration Stations,"
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 th e diesel engine to trip on overspeed.
A special prototypic test conducted at the La Salle facility verified the hardware and load aspects of the HPCS power supply concept. This test is described in topical report NEDO-10905, Revision 3.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 consiste nt with the recommendations of Regulatory Guide 1.9 Revision 3 as described in the Bases for Technical Specif ication B 3.8.1. Compliance with Regulatory Guide 1.9 Rev. 0, as an acceptable basis for the selecti on 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
.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 an alysis, design, and/or equipm ent 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 prev ent 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-12 Regulatory Guide 1.20, Revision 2, May 1976 Comprehensive Vibration Asse ssment Program for Reactor In ternals During Preoperational and Initial Startup Testing
Regulatory Guide Intent
Regulatory Guide 1.20 describe s a comprehensive vibration assessment program for reactor internals during pre operational and initial startup testing. The vibration assessment program meets th e 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 Alternat e 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 gui de through the incorporation of the alternate approach cited.
General or Alternat e Approach Statement
This regulatory guide is app licable to the core support structures and other reactor internals.
A vibration measurement program has been defi ned for the confirmatory testing of this plant during initial startup tests.
CGS reactor internals were test ed in accordance with prov isions 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-13 Regulatory Guide 1.21, Revision 1, June 1974 Measuring, Evaluating, and Re porting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liqui d and Gaseous Efflue nts from Light-Water-Cooled Nuclear Power Plants
Regulatory Guide Intent
Regulatory Guide 1.21 descri bes programs for measuring, reporting, and evaluating releases of radioactive mate rials in liquid and gaseous effluents and guidelines for classifying and reporting the categorie s and curie content of solid wastes.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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 monito ring and sampling system is designed to provide the monitoring and sampling capability required to make the measurements, evaluations, and reports r ecommended by this guide.
The radiation monitoring systems (RMS) provided to meet these objectives are
- a. For gaseous effluent streams Reactor building ventilati on exhaust plenum RMS
- b. For liquid effluent streams
- 1. Radwaste effl uent RMS, and 2. Service water RMS
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-14 c. For gaseous process streams
- 1. Offgas pretreatment RMS,
- 2. Offgas posttreatment RMS, and
- d. For liquid process streams
- 1. RHR service water RMS, and 2. Reactor building closed cooling water RMS These systems have th e 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 Specificat ions thereby satisfyi ng 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-15 Regulatory Guide 1.22, Re vision 0, February 1972 Periodic Testing of Protecti on 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 suppl y 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 COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-05-009 1.8-16 Specific Evaluation Reference
See above.
Similar Application Reference
Similar application has not b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-17 Regulatory Guide 1.26, Re vision 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 descri bes a quality classification system for determining acceptable quality standards fo r safety-related components containing water, steam, or radioactive material other than thos e components addresse d in 10 CFR 50.55a.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent used in this facility is in compliance with the intent of the subject regulator y guide through the incorporation of the alternate approach cited.
General Compliance or Alternate Approach Assessment
The definition of quality group classificati ons 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 opera tions and modifications comm ensurate with the safety functions performed by the safety-related components.
This regulatory guide is applicable to Quality Groups B thro ugh D pressure parts including piping, pumps, valv es, and vessels. Section 3.2 shows the quality groups classifications of these parts. The safety-related RCPB of the RWCU system meets the quality grouping requirements of Regulatory Guide 1.26. Non-safety-related portions of the RWCU system are maintained as Quality Group D vice C as delineated by Regulatory Guide 1.26.
Specific Evaluation Reference
See Section 3.2 and the Operational Quality Assurance Program Description (OQAPD).
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-06-000 1.8-18 Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 Require ments (Design and Construction).
Regulatory Guide Intent
- This guide describes an accep table method of complying wi th the NRC's regulations with regard to overall quality assurance program requirements.
Application Assessment
Assessed capability in design.
Compliance or Alternat e Approach Statement
The identified BWR Quality Assurance Program us ed 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 standar ds, 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 progra ms, are generally implemented across the board on all active proj ects with no opportunity for retrofit of completed work; therefore, wo rk performed later in a proj ect is typically subject to more quality assurance effort as a result of additional re quirements. Attachment B gives a graphic representation of the time relation of some of the major project activities with the date of issue of regu latory guides and the GE commitment dates.
Because of the long generation cycle of th e related ANSI Standard, GE had already COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 es tablish confidence that GE has been following a QA program which has kept current with customer and regulator y requirements.
Obviously, where most equipm ent 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 requi rements 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-06-014 1.8-21 Regulatory Guide 1.29, Re vision 3, September 1978 Seismic Design Classification Regulatory Guide Intent
- Regulatory Guide 1.29 describe s an acceptable method of id entifying 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 an alysis, design, and/or equipm ent 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 Seismi c Category I since the pumps do no t perform a safety function.
Specific Evaluation Reference
- See Section 3.2, Table 3.2-1
, and the OQAPD.
Similar Application Reference
- Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-22 Regulatory Guide 1.30, Revision 0, August 1972 Quality Assurance Requirements for the In stallation, Inspecti on, and Testing of Instrumentation and Electric Equipment.
Regulatory Guide Intent
This guide describes an acc eptable method of complying with the NRC's 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 re gulations 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 an alysis, design, and/or equipm ent 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 fe rrite is considered adequate to prevent microfissuring in austenitic stainless steel welds.
An extensive test program performed by G E, with the concurrence of the NRC, has demonstrated that controlling weld filler metal ferrite at 5% minimum produces
production welds which meet the requi rements 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-24 Regulatory Guide 1.32, Revision 1, March 1976 Use of IEEE 308-1974, "Criteria for Class 1E Electric System s for Nuclear Power Generating Stations"
Regulatory Guide Intent
This guide describes a method for implementati on of electrical safe ty related equipment design relative to GDC 17 a nd 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 a pplicable 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 an alysis, design, and/or equipm ent 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-25 Regulatory Guide 1.34, Re vision 0, December 1972 Control of Electrosla g Weld Properties.
Regulatory Guide Intent
Regulatory Guide 1.34 describes an acceptable method of implementing requirements regarding control of weld properties when fabricating electrosla g 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 regulat ory guide is not applicable.
Specific Evaluation Reference
Not applicable.
Similar Application Reference
Not applicable.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 acc eptable method of complying with the NRC's 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-27 Regulatory Guide 1.38, Revision 2, May 1977 Quality Assurance Requirements for Packaging, Shipping, Receiv ing, Storage and Handling of Items for Water-Cooled Nuclear Power Plants.
Regulatory Guide Intent
This guide describes an acc eptable method of complying with the NRC's requirements for handling of nuclear materials.
Application Assessment
- Assessed capability in design.
Compliance or Alternate Approach Statement
The identified BWR Quality Assu rance Program used in this facility reflec ts 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-28 Regulatory Guide 1.41, Revision 0, March 1973 Preoperational Testing of Redunda nt On-Site Electric Power Syst ems to Verify Proper Load Group Assignments.
Regulatory Guide Intent
The requirements of this regul atory guide are applicable to the total onsite electric power systems within Energy Northwest's responsibility.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent used in this facility is in full complia nce 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 cl adding on coarse grain low-alloy steel for safety cl ass components is not used.
General Compliance or Alternate Approach Assessment:
Not applicable.
Specific Evaluation
Reference:
Not applicable.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 pro cessing of stainless steel to avoid severe sensitization c ould 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 austen itic stainless steels.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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 specifie d by A262 A or E (Detecting Susceptibility to Intergranular Attack in Stainless Steel) dete cts sensitization in a gross way, and the tests do not provide a precise method of predicting susceptibility to stre ss 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 ra pid 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-31 These controls were used to avoid severe sensitiza tion and to comply with the intent of Regulatory Guide 1.44.
Specific Evaluation Reference
See Section 5.2.3. Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT December 2011 LDCN-11-005 1.8-32 Regulatory Guide 1.45, Revision 0, May 1973
Reactor Coolant Pressure Bounda ry 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 an alysis, design, and/or equipm ent 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 (i n 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 ca tegories 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.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-020 1.8-33 Large leaks are also detected by changes in reactor wa ter level and changes in flow rates in process lines.
The 5 gpm leakage rate is the limit on unidentified le akage inside the dr ywell. The leak detection system is capable of monitoring the flow rates with an accuracy of 1 gpm and is thus in compliance with paragr aph 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. Th is 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 ope rators to satisfy remainder of C.7.
The leakage detection systems are equipped with provisi ons to permit testing for operability and calibration during operation by the following methods:
- a. Continuous monitoring of sump leve l compared to flow rates into sump, b. Operability checked by compar ing 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.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT December 2011 LDCN-11-005 1.8-34 Limiting conditions for identif ied 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-35 Regulatory Guide 1.46, Revision 0, May 1973 Protection Against Pipe Wh ip Inside Containment
Regulatory Guide Intent:
Regulatory Guide 1.46 describes an acceptable basis for se lecting 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 an alysis, design, and/or equipm ent 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 compartmenta lization 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 ba sed on the pipe break evaluation.
Pipe whip requirements for fluid system pi ping within the primary containment that, under normal operation, has se rvice temperature greater than 200°F or pressures greater than 275 psig, complie d with ANS N176, "Design Ba sis for Protection Against Pipe Whip," and Regulatory Guide 1.46 excep t 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 lo ad sets (including zero load set) according to NB-3600 for upset and on op erating basis earthquake (OBE) event transient;
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-36 b. If Equation 10 resu lts in 2.4<S<3.0 S m for ferritic or aust enitic steels, the cumulative usage factor, U, calculat ed 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 Sm. Specific Evaluation Reference
See Section 3.6. Similar Application Reference
Similar application wa s used in GESSAR.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-37 Regulatory Guide 1.47, Revision 0, May 1973 Bypassed and Inoperable Status Indication for Nuclear Powe r Plant Safety Systems.
Regulatory Guide Intent
This guide describes an acc eptable 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 an alysis, design, and/or equipm ent 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:
Each safety-related system described in Sections 7.2, 7.3, 7.4, and 7.6 is provided with an automatically or operator initiated system level bypass and inoperability annunciator.
The system level annunciators are located with th e associated system controls and indications on main co ntrol room panels.
In addition to system level annunciation, co mponent and channel le vel annunciators are provided on other panels either in the control room near system controls or locally near affected equipment, to indicate the cause of the system bypass or inoperability.
A switch is provided for manua l actuation of each system level annunciator to allow display of those bypass or inoperable conditi ons which are not automatically indicated.
Typically, the following bypasses or inoperabilitie s cause actuation of system level (and component level) annunciati on for the affected system:
- a. Pump motor breaker not in operate position,
- b. Loss of pump motor control power,
- c. Loss of motor-operated valv e control power/motive power,
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-38 d. Logic power failure,
- e. Logic in test,
- f. Position of remote manual valves which do not receive automatic alignment signals, and
- g. Bypass or test switches actuated.
Auxiliary supporting system inoperability or bypass resulting in the loss of other safety-related systems will caus e actuation of system level a nnunciators for the auxiliary supporting system as well as those safety-related sy stems affected.
Specific Evaluation Reference
- See Section 7.1.2.4. Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-39 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 opera tion, postulated accidents, and specified seismic events for the desi gn of the Seismic Category I fluid system components.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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 Gu ide 1.48, see the attached tabulation.
The design basis was representa tive of good industry practic es at the time of design, procurement, and manufacture and is show n to be in genera l agreement with requirements of Regulatory Guide 1.48, with the following clarifications:
- a. The probability of an OBE of the magn itude postulated for CGS is consistent with its classification as an emer gency event. However, for design conservatism, loads due to the OBE vibration motion have been included under upset conditions; loads due to the O BE vibratory motion plus associated transients, such as a turbin e trip, have been consider ed in the equipment design under emergency conditions consistent with the probability of the OBE occurrence; and
- b. The use of increased stress levels fo r Class 2 components is consistent with industry practice as specified in ASME Code Section III.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-40 Specific Evaluation Reference
- See Section 3.9.3. Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment59 FINAL SAFETY ANALYSIS REPORTDecember 2007 1.8-41COMPARISON 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-32235/ 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-32235/ 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-32252/ 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 COLUMBIA GENERATING STATION Amendment59 FINAL SAFETY ANALYSIS REPORTDecember 2007 1.8-42COMPARISON 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)]
COLUMBIA GENERATING STATION Amendment59 FINAL SAFETY ANALYSIS REPORTDecember 2007 1.8-43COMPARISON 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 (o r comparative column) co rrespond to the following:
aIn addition to compliance with the design limits specified, assu rance of operability under all de sign loading combinations shal l be in accordance with Section 3.9.3.2. bReferenced 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 component s under faulted condition). See Section 3.9.5.2. dNot 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 c ondition is selected based o n the probability of occurrence of the particular load combination. The comb ination 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 COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-07-044 1.8-44 DELETED Contents of Regulatory Guide 1.49, Re vision 1, December 1973
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 a nd 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 an alysis, design, and/or equipm ent 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 ve ssel. Other ferritic components in the RCPB are fabricat ed 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 en sure removal of hydrogen, or preheat was ma intained until postweld heat treatment. The minimum pr eheat and maximum interpass temperature were specified and monitored.
All welds were nondestructively examined by radiographic methods. In addition, a supplemental ultrasonic exam ination 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 an alysis, design, and/or equipm ent used in this facility is in full complia nce with this regulatory guide.
General Compliance or Alternate Approach Assessment:
Compliance is achieved by sp ecifying, designing, and c onstructing 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-Fa ilure Criterion to Nuclear Power Generating St ation 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 re dundant instrument channels, logic, and actuation circuits to ensure that the singl e 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 acc eptable method of complying with QA requirements for protective coatings.
Application Assessment
- Assessed capability in design.
Compliance or Alternate Approach Statement
The identified BWR Quality Assu rance Program used in this facility reflec ts 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 acc eptable method of implemen ting 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 an alysis, design, and/or equipm ent 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 syst em 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 c onductivity and chloride concentrations.
Operationally, the conductivity is limited because it can be continuously and reliably measured and gives an indication of abnor mal 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 accep table method of complying w ith the NRC's regulations on qualification of nuclear power plant inspection, examina tion and testi ng personnel.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
The identified BWR Quality Assu rance Program used in this facility reflec ts 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-50 Regulatory Guide 1.60, Re vision 1, December 1973 Design Response Spectra for Seismic Design of Nuclear Power Plants.
Regulatory Guide Intents
This guide delineates procedures for defi ning 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 an alysis, design, and/or equipm ent 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 struct ures were given in terms of response spectra based on data available on earthquake acceleration time history records which was acce pted 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-51 Regulatory Guide 1.61, Re vision 0, October 1973 Damping Values for Seismic Design of Nuclear Power Plants Regulatory Guide Intent
- This guide delineates damping values that s hould be applied to mo dal dynamic analysis of Seismic Category I elements.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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 analys is conform to the data available on this at the time the analysis was pe rformed which was the prac tice 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-52 Regulatory Guide 1.62, Re vision 0, October 1973 Manual Initiation of Protective Actions.
Regulatory Guide Intent
Regulatory Guide 1.62 requires that manual initiation of e ach protective action at the system level be provided, that such in itiation accomplishes all actions performed by automatic initiation, and that protective action at the system level go to completion once manually initiated. In addition, manua l initiation should be by switches readily accessible in the control room, and a minimu m 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 an alysis, design, and/or equipm ent used in this facility is in full complia nce 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
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 manu al and automatic initi ation of the above function is kept to a minimum through implementation of manual activation as close as possible to the final devices actuators (rel ays, scram contractor
) of the protection system. No failure in the manual, auto matic or common portions of the protection system will prevent initiation of a give n function by manual or automatic means.
Manual initiation of any of the above functions, once initia ted, goes to completion as required by IEEE 279-1 971, 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-54 Regulatory Guide 1.64, Revision 2, June 1976 Quality Assurance Requirements for th e Design of Nuclear Power Plants
Regulatory Guide Intent
This guide describes an acceptable method of comp lying with the NRC's 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 Assu rance Program used in this facility reflec ts 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-55 Regulatory Guide 1.65, Re vision 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 bolti ng for light-water-cooled reactors.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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/a lcohol or a nickel pow der 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 approv ed for use in practice are judged to ensure comparable material quality and, moreover, are considered adequate on the basis of compliance w ith the applicable requireme nts 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 COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-56 and washers per ASME SA-388 in both an ax ial and circumferentia l direction. Any indication greater than the indication from the applicable calibration feature is unacceptable. A distance-amplitude correction curve pe r NB-2585 is used for the longitudinal wave examination. Surface exam inations were perfor med on the studs and nuts after final heat treatment and threading, as specified in the Regulatory Guide, in accordance with NB-2583 of ASME C ode Section III, 1971 Edition through November 1971 Addenda.
In relationship to regulatory position C.2, GE practice allows exposure of stud bolting surfaces to high purity fill water; nuts and washers are st ored dry during refueling.
Specific Evaluation Reference
See Section 5.3.1.7. Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-57 Regulatory Guide 1.66, Re vision 0, October 1973 Nondestructive Examination of Tubular Products.
Regulatory Guide Intent
This guide describes a method of impl ementing requirements acceptable to NRC regarding nondestructive examination requirements of tubul ar products used in the RCPB. Applicable Assessment
- Assessed capability in design.
Compliance or Alternate Approach Statement:
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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 we re supplied in accordance w ith applicable ASTM/ASME material specifications. Thes e specifications require a hydr ostatic 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 requirement s 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, A ppendix B requirement s and ASME code requirements assured adequate control of quality for the products.
This regulatory guide was withdrawn on Se ptember 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 wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-58 Regulatory Guide 1.67, Re vision 0, October 1973 Installation of Overpressure Protection Devices
Regulatory Guide Intent:
This regulatory guide describes a method accep table to the NRC sta ff for implementing GDC 1 with regard to the design of piping for safety valve and relief valve stations which have open discharge systems with lim ited discharge pipes a nd 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 ther mal 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 ex cess pressure caused by the di fference of thermal expansion between the pipe and the wa ter contained between the c ontainment isolation valves.
Specific Evaluation Reference
See Section 3.9.3.1.14
. Similar Application Reference
- Not applicable.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-59 Regulatory Guide 1.68, R evision 0, November 1973 Preoperational and Initial St artup Test Programs for Wa ter-Cooled Power Reactors
Regulatory Guide Intent
Regulatory Guide 1.68 describes the requirements for the initial startup test programs.
This regulatory guide is app licable to such activities as precritical tests and low-power tests. Application Assessment
- Assessed capability in design.
Compliance or Alternat e 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 a lternate acceptable appr oaches for specific conformance to this regulatory guide.
The format of the CGS test procedures is d ifferent from that of the guide, but since the content specifies the required elements, the procedur es 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 protec tion of public health and safety are included in the Technical Spec ifications for the plant. Th e General Electric startup instructions contain notes of caution which supplement the Techni cal Specifications.
The Technical Specifications should be the instrument for describing operational (including testing) limitations. Therefore, the identificati on 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
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-60 simulation tests are not possible. The guide wording would change to "... less than full simulation should be prov ided or referenced fo r 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 requi red knowledge of effective overall rod worth. Individual control r od calibrations cannot be performed in a meaningful manner in a large multirodded BWR. Therefore, th is 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 t han 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
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 pre operational test program.
Appendix A, Section D.2.b: Friction tests are performe d 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 appropr iate to limit the test to 50% and 100% power levels.
Appendix A, Section D.2.g
- At least six chemical anal yses 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 do es not include an emergency condenser, this secti on is not appropriate.
Appendix A, Section D.2.n: Control rod calibration in a large multirodded BWR has not been found to provide me aningful data. Any safety-r elated 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 st eam valve function tests are conducted at a minimum of six power and flow conditions
, the limitations of 25%, 50%, and 75% are not appropriate.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 act uation. 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 mini mum 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. Func tional 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, "demons tration of water injection fo r a loss-of-coolant accident" is not within the scope of the re actor 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 s hould be performed to check recirculation pump runback.
Appendix C, Section B.1.b
- Poison curtains are not app licable 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 in volve the approach of any safety margin or operating limit. The rod pa ttern exchange procedure at pow er is part of the Nuclear Performance Evaluation Procedure and will be performed during the fuel cycle as necessary. The simultaneous tr ip of both recirculation pump s 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 th erefore, it does not involve the approach of any safety margin or operating limit.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT December 2011 LDCN-10-029 1.8-63 Regulatory Guide 1.70, Re vision 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 an alysis, design, and/or equipm ent used in this facility is in full compliance wi th 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-64 Regulatory Guide 1.71, Re vision 0, December 1973 Welder Qualification for Ar eas of Limited Accessibility
Regulatory Guide Intent
Regulatory Guide 1.71 requires th at weld fabrication and re pair for wrought low-alloy and high-alloy steels or other materials such as static a nd centrifugal castings and bimetallic joints should comply with fa brication requirement s 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 an alysis, design, and/or equipm ent 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 rest rictive welds involved in the fabrication of BWR components. We lder qualification for welds with the most restrictive access was accomp lished 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 appr opriate AWS requirements.
Specific Evaluation Reference
See Section 5.2.3. Similar Application Reference
Similar application was used for Zimmer and LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-65 Regulatory Guide 1.73, Re vision 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 38 2-1972, "Trial-Use Guide for Type Test of Class 1 Electric Valv e Operators for Nuclear Power Generating Station." Regulatory position s tipulations are also included.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent used in this facility is in full complia nce with this regulatory guide.
General Compliance or Alternate Approach Assessment
This regulatory guide is applicable to the r ecirculation 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 expos ed to during and following a DBA (LOCA).
Specific Evaluation Reference
See Section 3.11. Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-66 Regulatory Guide 1.74, Re vision 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 Assu rance Program used in this facility reflec ts 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-67 Regulatory Guide 1.75, Re vision 0, February 1974 Physical Independence of Electrical Systems
Regulatory Guide Intent
This guide presents a detaile d method of ensuring physi cal independence of electric systems, including requirements of prepar ation, identificatio n, and isolation.
Application Assessment
Assessed capability in design
Compliance or Alternate Approach Statement
Identified NSSS scope of supply an alysis, design, and/or equipm ent 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 pr ior to their issuance.
The following is a point-by-point definiti on of the implementation of IEEE-384 as modified by Regulatory Guid e 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-68 4. General Separation Criteria 4.1 Required Separation 4.2 Equipment and Circu its Requiring Separation The equipment and circuits requiri ng separation are determined and delineated early in the plant design.
Distinctive identif ication 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 divi sional devices and wiring. Safety system logic is implemented with relay coil to relay cont act separation of multidivisional and nondivisional si gnals. 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-Cla ss 1E circuits a nd are separated to the extent that good elect rical isolation is assure
- d. This assurance was provided without Class 1E isolator
- s. Some physica l 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT 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 theref ore comply on that basis.
5.2 Standby Power Supply Comply as applied to the Divi sion 3 HPCS Diesel Generator.
- 5.4 Distribution System Comply as applied to the Divi sion 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 f Class 1E equipment and circuits are located on separate control switchboards or wher e operationally necessary on a single control switchboard.
- Division 1 and 2 power complia nce is provided in Section 1.8.3. f The control room structure and location as well as local control switchboard location is discussed in Section 1.8.3.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-70 5.6.2 Internal Separation Most of the devices requiri ng separation are separated by barriers. With several divi sions in one panel, and for relays which must accept mu ltidivisional signals, 6-inch separation is impossible. Ther efore, separation is done on a best effort approach. Desi gn 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 respectiv e 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.
COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 1.8-71 Specific Evaluation Reference
- See Section 8.3.1.4.2.7 Similar Application Reference
Application of this regulator y 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 ch anges to the ASME Code Cases and the current position of the NRC on acceptability for use. The guide cont ains tables that detail the NRC acceptance requi rements for current, annulle d, 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 applicat ions. The FSAR does not track individual Code Cases and revisi on 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 ha ve 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 contr actually 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 an alysis, design, and/or equipm ent 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 a nd 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 equipmen t 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-08-004 1.8-74 Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-75 Regulatory Guide 1.88, Re vision 2, October 1976 Collection, Storage, and Main tenance of Nuclear Power Plan t Quality Assura nce Records.
Regulatory Guide Intent
This guide describes an accep table method of complying with the NRC's regulations for collection, storage, and maintena nce of quality assurance records.
Application Assessment
Assessed capability in design.
Compliance or Alternate Approach Statement:
The identified BWR Quality Assu rance Program used in this facility reflec ts 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 b een used for other projects.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-020 1.8-76 Regulatory Guide 1.89, Revision 1, June 1984
Qualification of Class 1E Equi pment 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 Qualif ying Class 1E Equipment for Nuclear Power Generating Stations."
Additional regulatory positi on stipulations are also included.
Compliance or Alternate Approach Statement:
CGS complies with this re gulatory guide for equipment requiring environmental qualification procured after February 22, 1983.
General Compliance or Alternate Approach Assessment:
For equipment requiring enviro nmental qualification installe d prior to February 22, 1983, CGS follows the guida nce in NUREG-0588 Cat II.
In view of the NRC Memorandum and Order (CLI-80-21), date d 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 guid elines was found in specifi c equipment qualifications, additional testing and/or anal ysis was performed to demonstrate the adequacy of the equipment to perform its safety-related function.
Specific Evaluation
Reference:
Delineation of the degree of comp liance is contained in Section 3.11.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-77 Regulatory Guide 1.92, Re vision 1, February 1976 Combination of Modes and Spatial Components in Se ismic Response Analysis.
Regulatory Guide Intent
This guide describes methods acceptable to the NRC for combining the values of the response spectrum nodal dynamic an alysis 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 an alysis, design, and/or equipm ent 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 desi gn basis requirement of the construction permit for this plant.
The total seismic response is predicted by combining the response calculated from analys es due to one horizont al and one vertical seismic input. For this case, where the re sponse spectrum method of seismic analysis is used, the basis for combining the loads from the two an alyses is given as follows:
- a. The peak of the different modes fo r 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 m odes and due to differe nt excitations may not occur at the same location nor in the same direction.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-78 To implement the above, the two translation components of earthquake excitations are combined by summing the abso lute sum of all responses of interest (e.g., strain, displacement stress, moment, shear, etc.) fr om 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, stress es, 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 a bove method adequately demons trates 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 an alysis 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 respons es is defined mathematically as RRiin21 where R = Combined response
Ri = 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-79 Specific Evaluation Reference
- See Sections 3.7.3.6 and 3.7.3.7. Similar Application Reference
Similar application wa s used for LaSalle.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 th e prediction of irradi ation 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/tempe rature limit curves are in full compliance with the identified requireme nts 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 fo r the limiting belt line material for the vessel. Using linear elastic fracture mechanics, the requi rements of Welding Research Council Bulletin 175, the Standard Review Plan, and the require ments of Regulatory Guide 1.99, Revision 2, the pressure/temperature limit curves were developed for CGS. These curves will be used to ev aluate the predictions determined by the regulatory guide until the submittal of new curv es 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 e ndorses both the requirement s and recommendations of IEEE 344-1975, "IEEE Recommended Practices fo r Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations," when such qualification is performed in conjunction with Regulatory Guide 1.89, a nd 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 th e 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 lo ad combinations.
Specific Evaluation Reference
See Section 3.10 and "WNP-2 Dynamic Qualification Report for Safety-Related Equipment," dated September 1982
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 an alysis, 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 th e 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-83 Regulatory Guide 1.183, Revision 0, July 2000 Alternative Radiological Source Terms For Ev aluating Design Basis Ac cidents 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 eval uations; consideration of im pacts on analyzed risk; and content of submittals. This guide establishes an accepta ble alternativ e source term (AST) and identifies the significant attri butes of other ASTs that may be found acceptable by the NRC staff. This guide also identifies acceptable radiological analysis assumptions for use in conjunc tion with the accepted AST.
Application Assessment
- Assessed capability in design.
Compliance or Alternate Approach Statement
- Identified NSSS scope of supply an alysis, 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-FT S-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 regul atory guide lists the elements in each radionuclide group that should be considered in design basis an alyses. The intent of the guidance is met by an alternat e approach. The Co lumbia analyses consider 66 nuclides consisting of 60 identified as being potentially important contributors to TEDE in
NUREG/CR-4691 plus seve n 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-84 [Guide Section 3.3 of A ppendix 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 re moved 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 pa rticulate 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-85 Regulatory Guide 1.190, Revision 0, March 2001 Calculational and Dosimetry Methods for Determining Pressu re 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 st aff 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 genera l, the methodology described in the LTR adhere s to the guidance in Regulatory Guide 1.190 for neutron flux evaluation and was approve d by the U.S. NRC in th e 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, Un its 2 and 3, reactor vessels.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-86 Regulatory Guide 1.194, Revision 0, June 2003 Atmospheric Relative Concentra tions for Control Room Radiol ogical 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 ro om radiological habita bility assessments at nuclear power plants. This guide describes methods accep table to the NRC staff for determining /Q values that will be used in cont rol 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 an alysis, 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
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-87 1.8.3 BALANCE OF PLANT SC OPE OF SUPPLY EVALUATION The following evaluations of implementation of re gulatory guides are relati ve 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 implementa tion 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 guida nce 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 a nd 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 thei r revisions are addressed in the following.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-88 Regulatory Guide 1.6, Revision 0, March 1971 Independence Between Redundant Standby (Onsite) Power S ources 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 assessm ents 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 load s, 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 st andby 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 auto matic connection to any other redundant dc load group.
- 4. When operating from the standby sources, redundant load groups and the redundant standby sources are i ndependent of each other.
- 5. A single generator driven by two prim e 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 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 minimum educational and experience qualifications for the onsite plant personnel with the exception of the Health Physics/Chemistry Supervisor are based on ANSI 18.1-1971, "Standard fo r Selection and Training of Personnel for Nuclear Power Plants," which is referenced by Regulatory Guide 1.8.
Qualification requirements for the Health Physics/Chemistry Supervis or are as set forth in this guide.
Specific Evaluation Reference
See Sections 13.1.3, 13.2.1, and the OQAPD.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT 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 assessm ents 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 Di vision 2 diesel generator sets were selected to have a continuous load rating equal to or great er 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 Divisi on 2 diesel generator sets are capable of starting and accelerating to rated speed, in the required sequence, all the needed engineered safety feature and emer gency shutdown loads.
The Division 1 and Division 2 diesel generator sets are within the limits of undervoltage, under-frequency, overspeed and voltage a nd frequency restoration time limits, set forth in the regulatory guide.
- 5. The suitability of each diesel generator set of the standby powe r supply was confirmed by prototype qualification test data a nd 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 consiste nt with the recommendations of Regulatory Guide 1.9 Revision 3 as described in the Bases for Technical Specif ication B 3.8.1. Compliance with Regulatory Guide 1.9 Rev. 0, as an acceptable basis for the selecti on 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-91 Regulatory Guide 1.10, Re vision 1, January 1973 Mechanical (Cadweld) Splices in Reinforced Bars of Cate gory 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 incl uded 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 subs tandard 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 acceler ographs 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 st ructures, systems, and components.
Specific Evaluation Reference
- See Section 3.7.4.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-94 Regulatory Guide 1.13, Re vision 1, December 1975 Spent Fuel Storage Facility Design Basis
Compliance or Alternate Approach Statement
- CGS complies with the intent of the guidan ce set forth in this regulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment
- A controlled leakage building is provided en closing the fuel pool. The building is not designed to withstand extremely high winds, but leakage is suitably controlled during refueling operations. The build ing is equipped with a ventil ation 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, th e 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 interloc ks is permitted only dur ing 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-95 Regulatory Guide 1.15, Re vision 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 incl uded in the appropriate specifications for project construction. Complia nce with the guide is assured by the implementation of qualified testing and control procedures a nd 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-97 Regulatory Guide 1.17, Revision 1, June 1973 Protection of Nuclear Power Plan ts 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-98 Regulatory Guide 1.18, Re vision 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-08-000 1.8-100 Regulatory Guide 1.21, Revision 1, June 1974
Measuring, Evaluating, and Reporting of Radioactivity in So lid Wastes and Releases of Radioactive Materials in Liqui d and Gaseous Efflue nts 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 m onitoring 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 monito ring systems adequa tely monitor efflue nt discharge paths for radioactivity that may be released from normal operations, including anticipated operational occurrences, and fr om postulated accidents.
Columbia Generating Station complies w ith Section C.11.b (Quality Controls) requirements for blind duplicate analysis by an alternate a pproach. An intralaboratory blind sample program is performed on selected samples. The blinds are prepared from samples sent from a cross check laboratory and split betw een several analysts as determined by the Chemistry Supervisor or de signee. 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 a ppropriate standards be used to calibrate continuous radioactivity monitors and that the relationship be established between
monitor readings and concentration over the fu ll 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 calcula tions will be used. Subsequent inservice calibrations will be performed using the spec ific radionuclide analyt ical results from grab samples taken from the effluent release path.
Appendix A, Section A.3.a (1) and Section A.3.a (3), anal ytical frequencies are not consistent with standard sa mpling and analytical techniques. Improved sensitivities and COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-101 more realistic quantity measurements can be ma de by performing 140Ba-La, 89-90Sr, 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 en trained fission and activation gases. The gamma spectrum an alysis performed prior to the release of any waste liquid batch will identify such gases without perf orming a separate or special analysis.
The sensitivity slated in Appendix A, Se ction 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-102 Regulatory Guide 1.22, Re vision 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 react or 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-103 Regulatory Guide 1.23, Re vision 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 re commendations 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 regul atory guide for a meteorological program to provide the meteorological data required to estimate pot ential 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-104 Regulatory Guide 1.26, Re vision 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 a nd VIII. Quality group classifications have been ma intained during design and co nstruction. Quality group classifications are maintained during plant operations and modifications by plant administrative procedures and the plant m odification control pr ocess. The quality group classifications are commensurate with the safety functions performed by the safety-related components.
The turbine stop valves and by pass valve, which are classi fied 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 and the OQAPD.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-105 Regulatory Guide 1.27, Re vision 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 make up 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 cons ider two or more mo st severe natural phenomena occurring simultaneously, the TMU was designed to be tornado proof but was not designed and construc ted to withstand the effe cts 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-106 Regulatory Guide 1.28, Revision 0, June 1972 Quality Assurance Program Requirements (Design and Construction)
Compliance or Alternat e Approach Statement:
CGS complies with the guidance se t forth in this regulatory guide as described below.
General Compliance or Alternate Approach Assessment
- Procurement documents issued after Novem ber 1973 required compliance with ANSI N45.2. Prior to that time, an "expl anative 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 COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-107 Regulatory Guide 1.29, Re vision 3, September 1978 Seismic Design Classification
Compliance or Alternate Approach Statement
- CGS complies with the intent of the guidan ce 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. Th e 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 su ction 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-108 Regulatory Guide 1.30, Revision 0, August 1972 Quality Assurance Requirements for the In stallation, Inspecti on, and Testing of Instrumentation and El ectrical 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 activi ties performed, except in those isolated instances where requirements were entered direc tly 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-110 Regulatory Guide 1.32, Re vision 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 docke ted 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 t hose 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 th e 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-111 Regulatory Guide 1.33, Re vision 2, February 1978 Quality Assurance Program Requirements (Operation)
Compliance or Alternate Approach Statement:
CGS complies with the intent of the guidan ce set forth in this regulatory guide by an alternate approach.
Compliance or Alternat e Approach Assessment
- Compliance is discussed in the OQAPD.
Specific Evaluation Reference
- See Section 13.5.1.1 and the OQAPD.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-112 Regulatory Guide 1.34, Re vision 0, December 1972 Control of Electros lag 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-113 Regulatory Guide 1.35, Re vision 2, January 1976 Inservice Inspection of Ungrout ed Tendons in Prestressed Conc rete 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-114 Regulatory Guide 1.36, Re vision 0, February 1973 Nonmetallic Thermal Insulation for Austenitic St ainless 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 requ irements of this regulatory guide.
Specific Evaluation
Reference:
Not applicable.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 co mpliance 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-116 Regulatory Guide 1.38, Revision 2, May 1977 Quality Assurance Requirement fo r 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 th e intent of this guidance by an alternate approach.
The changes to the regulatory positions of Re vision 1 and 2 of this regulatory guide,
which specify additional detail ed requirements and make certa in nonmandatory sections of ANSI N45.2.2 mandatory
, are not implemented.
General Compliance or Alternate Approach Assessment
- Procurement documents required compliance wi th ANSI N45.2.2, Revision 0, and/or contained a generic specifi cation packaging section a nd/or specified directly requirements for these functions.
The regulatory positions contai ned in Revision 1 and 2 of this regulatory guide changed significantly from the origin al 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 regula tory 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-117 Regulatory Guide 1.39, Re vision 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 app licable housekeeping requirements.
Specific Evaluation Reference
- None II Operational Phase Compliance is discussed in the OQAPD.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-118 Regulatory Guide 1.40, Revision 0, March 1973 Qualification Tests of Conti nuous Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants
Compliance or Alternate Approach Statement
- CGS complies with the guidance se t forth in the regulatory guide.
General Compliance or Alternate Approach Assessment
- Containment fans have been qualified for in containmen t use in accordance with IEEE 334-1974.
Specific Evaluation Reference
- See Section 9.4.11.3.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-119 Regulatory Guide 1.41, Revision 0, March 1973 Preoperational Testing of Redunda nt On-Site Electrical Power Sy stems to Verify Proper Load Group Assignments
Compliance or Alternate Approach Statement
- CGS complies with the guidance se t forth in the regulatory guide.
General Compliance or Alternate Approach Assessment:
As part of the preoperationa l 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 resp ective load groups.
Specific Evaluation Reference
- See Sections 8.1.5.2, 8.3.1.2.2 and 14.2.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 safe ty class components.
General Compliance or Alternate Approach Assessment
- Not applicable.
Specific Evaluation Reference
- Not applicable.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 weldin g 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 fo rth 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.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT December 2011 LDCN-11-005 1.8-122 Regulatory Guide 1.46, Revision 0, May 1973
Protection Against Pipe Wh ip Inside Containment
Compliance or Alternate Approach Statement:
CGS complies with the intent of the guidan ce 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 Tec hnical 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.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT December 2011 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 guidan ce set forth in this regulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment:
Each safety-related system described in Sections 7.2, 7.3, 7.4, and 7.6 is provided with an automatically or operator initiated system level bypass and inoperability annunciator.
The system level annunciators are located with th e associated system controls and indications on main co ntrol room panels.
In addition to system level annunciation, co mponent and channel le vel annunciators are provided on other panels either in the control room near system controls or locally near affected equipment, to indicate the cause of the system bypass or inoperability.
A switch is provided for manua l actuation of each system level annunciator to allow display of those bypass or inoperable conditi ons which are not automatically indicated.
Typically, the following bypasses or inoperabilitie s cause actuation of system level (and component level) a nnunciation for the af fected systems:
- a. Pump motor breaker not in operate position,
- b. Loss of pump motor control power,
- c. Loss of motor-operated valv e control power/motive power,
- d. Logic power failure,
- e. Logic in test,
- f. Position of remote manual valves which do not receive automatic alignment signals, and
- g. Bypass or test switches actuated.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT December 2011 1.8-124 Auxiliary supporting system inoperability or bypass resulting in the loss of other safety-related systems will caus e actuation of system level a nnunciators for the auxiliary supporting system as well as those sa fety-related syst ems affected.
Specific Evaluation
Reference:
Not applicable.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT 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
.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT 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 wh ich 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.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT 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 l onger applicable.
General Compliance or Alternate Approach Assessment:
Inservice inspection of CGS is based on ASME Section XI for Clas ses 1, 2, and 3.
Specific Evaluation
Reference:
See Section 3.9.6.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT December 2011 1.8-128 Regulatory Guide 1.52, Revision 2, March 1978 Design, Testing, and Maintena nce Criteria for Atmosphere Cl eanup 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 guidan ce given in Revision 2 of this regulatory guide.
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 po sition of Regulatory Guide 1.52, Revision 2, Ar ticle 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
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT 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 Criteri a 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
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT 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 Cr iteria 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 sta ndby gas treatment system are located in secondary containment and are not subject to containment pressure surges during accidents.
Redundant Seis mic 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.
COLUMBIA GENERATING STATION Amendment 61 FINAL SAFETY ANALYSIS REPORT 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 pre ssure 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 reacto r 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 pre ssure 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 in stalled demisters will protect the HEPA filters in the system fr om blinding unde r 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, th e FARR tests and qualification are considered adequate.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 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 fi lters 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 un its 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-133 Note 12 The laboratory testing crite ria 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 Syst em 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 interpre t with the 'U' sh aped charcoal beds installed due to air fl ow disturbance caused by the measuring apparatus. This is particularly true on the parallel legs of the 'U' shaped beds, where th e 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 pe rformed in accordance with ASME N510-1989, Sections 10 and 11, respectively.
Specific Evaluation Reference
- See Section 6.5.1.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 se t forth in the regulatory guide.
General Compliance or Alternate Approach Assessment
- Regulatory Guide 1.53 provide s guidance for the applica tion 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 c onformance with IEEE 379-1972 and the four supplemental conditions not ed in Regulatory Position C.
Specific Evaluation Reference
- See Section 8.1.5.2.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 coa tings in primary containment areas are manufactured and applied in accordance with quality assurance requirem ents of ANSI N101.4.
Specific Evaluation Reference
- See Section 6.1.2.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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. Co mpliance with the guide is a ssured by the application of appropriate concrete specif ications, construction practic es, codes and standards, including the documents recommended by the guide, for the placement of concrete; by the implementation of approved communications procedures be tween qualified design and construction forces; and by implementation of an approved QA program which
ensures design control and coor dinated quality control of concrete material, placement, inspection and testing between app licant, 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 guidan ce 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 reco rding of the conde nser 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 demine ralizer unit and fr om 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 de mineralizer system, and
- g. Excess condensate deminera lizer capacity to permit recharging of resin beds during normal plant operation.
Specific Evaluation Reference
- See Section 5.2.3.2.2
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 guidan ce of the regulatory guide by providing the following:
- a. Operating limits are prescribed for condensate filter demi neralizers. Plant operating conductivity limits are define d for the RWCU demineralizers. Effluent conductivity for the individual de mineralizers is recorded and a main control room alarm is triggered when conductivity limits are reached or exceeded;
- b. Condensate filter demine ralizer conductivity and flow instrumentation are used in the general assessment of individual demineralizer un it performance and capacity;
- c. An operational limit is set for hotwell c onductivity which triggers a main control room alarm. Hotwell conductivity, in conjunction with precalculated assessment of condenser inleakage rate s and demineralizer perfor mance permits appropriate action to be taken on exceedi ng the operating limit setpoint;
- d. Laboratory analyses ar e performed for chloride, pH, and conductivity at intervals appropriate to the plant ope rating 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 inco rporated into the CGS design. The general guidance of this item will, however, be applied to the pressure precoat filter demineralizer system
- s. Each lot of pr ecoat resins will be analyzed for capacity and im purity 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 guidan ce 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 accide nt recovery flooding plus OBE. For this load combination, the stress limits used for CGS are within the limits set forth in the NRC Standard Revi ew 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 Isla nd Standard Design dated December 1975.
Specific Evaluation Reference
- See Section 3.8.2.3.10
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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. Docume nted evidence of the above tr aining was maintained. The indoctrination and training comp lied with the requirements of Appendix B
, 10 CFR Part 50, and ANSI N45.2.
As of November 1980, in addition to the indoctrination and trai ning requirements noted above, requirements which meet this regulatory guide were imposed on site contractors for personnel performing inspections, examin ations, 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. Theref ore, specific capa bility 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-142 Regulatory Guide 1.60, Re vision 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 guidan ce set forth in this regulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment
- CGS meets the seismic requireme nts previously acceptable to the NRC as discussed in Section 3.7.1.1. Specific Evaluation Reference
- See Section 3.7.1.1.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-143 Regulatory Guide 1.61, Re vision 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 fo r CGS. The more conservative CGS design satisfies the requirements of Regulatory Guide 1.61.
Specific Evaluation Reference
- See Section 3.7.1.3.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-144 Regulatory Guide 1.62, Re vision 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 contro l 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-145 Regulatory Guide 1.63, Revision 2, July 1978, and Revision 3, February 1987 Electric Penetration Assemblie s 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 applicat ions docketed after August 31, 1978 and Fe bruary 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 as sessment given below correspond num erically to the regulatory positions as indicated in Section C of Regulatory Guide 1.63, Revision 0, October 1973.
- 1. Capability of with standing maximum fault I 2T heating in the case that overload protective devices fail:
CGS is in compliance with this require ment. 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 I2T heating experienced by the penetrati on conductors to levels below the conductor ratings.
- 2. The maximum containment pressure specified for CGS complie s with the safety margins required by the ASME B&PV Code, Article N3000, footnote 1.
- 3. The position refers to sp ecific applicability or acceptability of other codes, standards, and guides covered sepa rately 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-146 Regulatory Guide 1.64, Revision 2, June 1976 Quality Assurance Requirements for th e Design of Nuclear Power Plants I. Design and Construction Phase
Compliance or Alternat e Approach Assessment
- Regulatory Guide 1.64, Revisi on 0, Revision 1, and Revision 2 do not apply to CGS since they apply to construction pe rmits 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-147 Regulatory Guide 1.67, Re vision 0, October 1973 Installation of Overpressure Protection Devices
Compliance or Alternate Approach Statement
- This regulatory guide is not applicable to CG S 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-148 Regulatory Guide 1.68, Revision 1, January 1977 Initial Test Programs for Water-Cooled Reactor Power Plants
Compliance or Alternat e Approach Statement
- This regulatory guide is not applicable to the CGS initial te st 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 tes ting 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 Regulat ory Guide 1.68, Revision 0.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 guidan ce set forth in this regulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessments:
The preoperational testing and the initial St artup testing as described in Section 14.2 complies with the intent of this regulatory guide. Howeve r, due to the limitations of the auxiliary steam supply system, the conf irmation that the feedwater pumps satisfy required head, flow rate and suction head will not occur un til 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-150 Regulatory Guide 1.68.2, Revision 0, January 1977 Initial Startup Test Program To Demonstrate Remote Shutdown Capab ility For Water-Cooled Nuclear Power Plants
Compliance or Alternate Approach Statement
- CGS complies with the intent of the guidan ce 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 initiate d by scramming plant from th e control room versus a location outside the control room as desc ribed in Section C.3 of the regulatory guide. This exception is made to bette r simulate the actual procedure which would be followed if a control evacuati on 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 proce dure as described in Section C.4 of the Regulatory Guide may not be perf ormed immediately following the
demonstration of achieving and maintain ing 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 provisi ons for a delay in the demonstration of cooldown.
Specific Evaluation Reference
See Sections 14.2.12.3.28 and 7.4.1.4.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-151 Regulatory Guide 1.69, Re vision 0, December 1973 Concrete Radiation Shields for Nuclear Power Plants
Compliance or Alternate Approach Statement
- CGS complies with the intent of the guidan ce set forth in this regulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment
- Although the regulatory guide was prom ulgated after desi gn and specification implementation of the engineering criteria, the recommended design and construction practices specified in the re gulatory 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-152 Regulatory Guide 1.70, Re vision 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 require ments of Regulatory Guide 1.70, Revision 2. This includes both format and content.
Specific Evaluation Reference
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-153 Regulatory Guide 1.71, Re vision 0, December 1973 Welder Qualifications for Ar eas of Limited Accessibility
Compliance or Alternate Approach Statement
- CGS complies with the intent of the guidan ce 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 we ld 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 ex tensions to the torch handle or electrode holder, the contractor notifies the weldi ng engineer. All lim ited 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 qualifi cation test that simulates the limited access condition is required by th e welding engineer. For welds in the pressure retaining components the welder's 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-154 Regulatory Guide 1.72, Re vision 0, December 1973 Spray Pond Plastic Piping
Compliance or Alternate Approach Statement
in its spray ponds.
General Compliance or Alternate Approach Assessment
- Not applicable.
Specific Evaluation Reference
- Not applicable.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-155 Regulatory Guide 1.73, Re vision 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 duri ng and after a design basis accident are used in the tests to assure that the testing is carried out u nder more severe environmental conditions than those expected.
Specific Evaluation Reference
- See Sections 3.11 and 8.1.5.2.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-156 Regulatory Guide 1.74, Re vision 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 cl arified at the point of application.
Specific Evaluation Reference
- None II Operational Phase Compliance is discussed in the OQAPD.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-157 Regulatory Guide 1.75, Re vision 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 applicat ions docketed afte r February 1974. However, CGS complies with the intent of th e guidance set forth in this re gulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment
- See Section 8.3.1.4.2.7 for an assessment of CGS re lative to this regulatory guide.
Specific Evaluation Reference
- See Section 8.3.1.4.2.7
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 guidan ce set forth in this regulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment
- The tornado design criteria fo r Columbia Generating Sta tion were revised based on design basis tornado charact eristics 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 cr iteria were found acceptable by the NRC.
Specific Evaluation Reference
- See Section 3.3.2.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-159 Regulatory Guide 1.78, Revision 0, June 1974 Assumptions for Evaluating the Ha bitability 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 t oxicity limits presented in Re vision 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-160 Regulatory Guide 1.80, Revision 0, June 1974 Preoperational Testing of Instrument Air Systems
Compliance or Alternat e Approach Statement
- CGS complies with the guidance se t forth in this regulatory guide.
General Compliance or Alternate Approach Assessment
- The primary containment instrument air system preoperational test procedure incorporated the requirement s of this regulatory guide.
Specific Evaluation Reference
- See Sections 14.2.7.3 and 14.2.12.1.34
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-161 Regulatory Guide 1.82, Revision 0, June 1974 Sumps for Emergency Core Coolin g 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 ch anges to the ASME Code Cases and the current position of the NRC on acceptability for use. The guide cont ains tables that detail the NRC acceptance requi rements for current, annulle d, 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 applicat ions. The FSAR does not track individual Code Cases and revisi on 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 wr itten 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-164 Regulatory Guide 1.88, Revision 2 October 1976 Collection, Storage and Maintenance of Nuclear Power Plan t 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 regulato ry guide and construction of the CGS records facility, Project Qua lity Assurance complie d 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, storag e, and maintenance of quality assurance records by Project Quality A ssurance has been in complia nce with ANSI N45.2.9 and NFPA No. 232-1975 for fire pr otection as imposed by this regulatory guide. The record facility has a minimum of a 2-hr rating.
Procurement documents directly specify re quirements for collection, storage, and maintenance of records. Th e requirements generally meet the intent of ANSI N45.2.9 except that storage facilities or cabinets are only require d to meet a 1-hr rating.
II Operational Phase Compliance is discussed in the OQAPD.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-020 1.8-165 Regulatory Guide 1.89, Revision 1, June 1984
Qualification of Class 1E Equi pment for Nuclear Power Plants
Regulatory Guide Intent:
Regulatory Guide 1.89 endorses both the requirements and recommendations of IEEE 323-1974, "IEEE Standard for Qualif ying Class 1E Equipment for Nuclear Power Generating Stations."
Additional regulatory positi on stipulations are also included.
Compliance or Alternate Approach Statement:
CGS complies with this re gulatory guide for equipment requiring environmental qualification procured after February 22, 1983.
General Compliance or Alternate Approach Statement:
For equipment requiring envi ronmental qualification instal led prior to February 22, 1983, CGS follows the guida nce in NUREG-0588 Cat II.
In view of the NRC Memorandum and Order (CLI-80-21), date d 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 equi pment qualifications, additional testing and/or analysis is performed to demonstrate the adequacy of the equipment to perform its safety-r elated function.
For equipment whose qualification program has not been completed, a justification for interi m operation in accordance with 10 CFR 50.49 is performed as described in the "WNP-2 Environmen tal Qualification Report for Safety-Related Equipment," Reference 3.11-1.
Specific Evaluation
Reference:
See Section 3.11.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-166 Regulatory Guide 1.90, Revision 0, November 1974 In-Service Inspection of Prestr essed Concrete Containment St ructures with Grouted Tendons
Compliance or Alternate Approach Statement
- This regulatory guide is not applicable because CGS does not have a prestressed concrete containment struct ure with grouted tendons.
General Compliance or Alternate Approach Assessment
- Not applicable.
Specific Evaluation Reference
- Not applicable.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-167 Regulatory Guide 1.91, Re vision 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 applicati ons docketed on or af ter 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 th an the wind pressures caused by the design basis tornado. Therefor e, postulated explosi ons 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-168 Regulatory Guide 1.92, Re vision 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 19
- 76. 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 moda l responses has been implem ented in acco rdance with the guide's recommendations.
The combining of spatial components was perf ormed prior to the i ssuance 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 mo re 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-169 Regulatory Guide 1.93, Re vision 0, December 1974 Availability of Elec tric 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 doc keted after October 15, 1976. However, CGS complies with the intent of the guidan ce set forth in the guide.
General Compliance or Alternate Approach Assessment
- The guidelines included in ANSI 45.2.5-1974 fo r installation, inspec tion and testing of structural concrete and structural steel, in cluding nonpressure vess el 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-05-009,07-025 1.8-171 Regulatory Guide 1.95, Re vision 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 CG S since chlorine gas is not stored at CGS or nearby facilities and the exp ected quantities of chlorine sh ipped within five miles is less than the threshold volumes specified in Regulatory Guide 1.78.
Specific Evaluation Reference
- See Section 6.4.4.2.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-172 Regulatory Guide 1.97, Re vision 2, December 1980 Instrumentation for Light Water Cooled Nucl ear Power Plants to Assess Plant Conditions During and Following an Accident
Compliance or Alternate Approach Statement
- The CGS safety-related display instrume ntation meets the intent of Regulatory Guide 1.97.
General Compliance or Alternate Approach Assessment
- Instrumentation is provided in the main control room to monitor plant variables and systems during and following an accident. Th e instrumentation is qualified to remain functional as required by the regulatory guide.
Portable multichannel gamma-ray spectrometer instrumentation provided for use by field teams during emergencies is not used at CGS, c ontrary to the recommendation contained in Regulatory Guide 1.97, Revision 2, Tabl e 2, Plant and Environs Radioactivity (portable instrumentation).
Regulatory Guide 1.97, recommends the use of these instruments for rele ase assessment and analysis.
Alternative methods that produce more reliable indication of fuel fa ilure during a radioactive release are used instead, such as air sample analysis and validation of dose projections using field team sample results.
Specific Evaluation Reference
- See Section 7.5.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 e ndorses both the requirement s and recommendations of IEEE 344-1975, "IEEE Recommended Practices fo r Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations," when such qualification is performed in conjunction with Regulatory Guide 1.89, a nd 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-175 Regulatory Guide 1.102, Re vision 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 histor ical flood. Based on the crite ria stipulated in Regulatory Guide 1.102, the CGS plant site is classified as a "Dry Site."
Specific Evaluation Reference
- See Section 2.4.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-176 Regulatory Guide 1.103, Revision 1, October 1976 Post-Tensioned Prestressed Systems for Concrete Reactor Vessel s 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-177 Regulatory Guide 1.104, Re vision 0, February 1976 Overhead Crane Handling Systems for Nuclear Power Plants
Compliance or Alternate Approach Statement
- CGS complies with the intent of the guidan ce 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 spen t fuel cask handling mode, and
- d. Main hoist "paddle" type upper limit switch to preven t the inadvertent "two-blocking" condition.
Specific Evaluation
Reference:
See Sections 3.8.4.1.1.5 and 9.1.4.2.2
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 D ecember 15, 1976.
General Compliance or Alternate Approach Assessment
- Instrumentation is provided in a main cont rol room to monitor plant variables and systems. The range of instrumentation is selected to cover the anticipated ranges of variables for the follo wing 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-179 Regulatory Guide 1.106, Revision 1, March 1977 Thermal Overload Protection for Electri c 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 applicati ons docketed after July 15, 1976. However, CGS design complies with the intent of the guidance set forth in S ection C.2 of the regulatory guide.
General Compliance or Alternate Approach Assessment:
Class 1E motor-operated valve (MOV) overl oads are chosen tw o sizes above those which would be required base d 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-180 Regulatory Guide 1.107, Re vision 1, February 1977 Qualifications for Cement Gr outing 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.
COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-11-002 1.8-181 Regulatory Guide 1.108, Re vision 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 doc keted after April 1, 1977.
General Compliance or Alternate Approach Assessment:
Preoperational and periodic testi ng 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 consiste nt with the recommendations of Regulatory Guide 1.9 Revision 3 as described in the Bases for Technical Specif ication B 3.8.1. Regulat ory Guide 1.9 Revision 3 includes pertinent guidance for periodic testing previously a ddressed in Regulatory Guide 1.108. Specific Evaluation
Reference:
See Sections 8.3 and 14.2.12.1
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 re gulatory guide by usi ng Battelle Northwest models which are accep table to the NRC.
Specific Evaluation Reference
- See Sections 11.2.3.3, 11.3.3.3, and 5.2 of the Environmental Report.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-184 Regulatory Guide 1.111, Revision 1, July 1977 Method for Estimating At mospheric 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 guidan ce set forth in this regulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment
- Analyses of atmospheric tr ansport and dispersion of ga seous effluents at CGS are performed using the standard NRC diffusion models in NUREG/CR-2919, XOQ/DOQ: Computer Program for the Meteorological, Ev aluation of Routine Eff luent Releases at Nuclear Power Stati ons, September 1982.
Specific Evaluation References
- See Section 2.3.5.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 rele ases 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 be low for discussions of the methods currently used.
Specific Evaluation Reference
- See Sections 11.2.3.2 and 11.3.3.3.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 re leases 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 a nd 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 S ections 5.1, 5.2.2, and 6.1.1.1.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-187 Regulatory Guide 1.114, Revision 1, November 1976 Guidance to Operator at the Cont rols 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 doc keted 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 fr om low trajectory missiles.
Specific Evaluation Reference
- See Section 3.5.1.3.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 guidan ce 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 qua lity assurance program in co mpliance with ANSI N45.2.
Specific Evaluation Reference
- None II Operational Phase Compliance is discussed in the OQAPD.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 Fe bruary 15, 1977.
General Compliance or Alternate Approach Assessment
- Essential systems are protected from tornadoes by structur es designed for design basis tornadoes (DBT). See Regulator y Guides 1.27 and 1.76.
Specific Evaluation Reference
- See Sections 3.3.2.4 and 9.2.5.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 applicati ons docketed after Fe bruary 28, 1977. However, the NRC requested a reevaluation of th e fire protection program of CGS and a comparison with the guidelines in Appendix A to Branch T echnical Position APCSB 9.5-1, "Guidelines for Fire Protection For Nuclear Power Plants, Dock eted Prior to July 1, 1976." CGS complies with the intent of the guidance set forth in Appe ndix 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-193 Regulatory Guide 1.122, Re vision 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 guidan ce 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-194 Regulatory Guide 1.123, Revision 0, October 1976 Quality Assurance Requirements fo r 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 guidan ce 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 guid e, requires certain supplier selection, evaluation, and pre- and post-award activities.
Prequalification of s uppliers was generally not performed. The procurement documents required prospective suppliers to submit inform ation 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 evaluation could not be accomplished with the information supplied. Post-a ward evaluations were perfor med 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 Ener gy Northwest of all in spections and tests which were selectively witn essed by Energy Northwest.
Specific Evaluation Reference
- None II Operational Phase Compliance is discussed in the OQAPD.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 th ose 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 appr opriate combinations of loadings which reflect the re quirements set forth in the 197 4 Edition of the ASME Code Section III, Subsection NF, along with additi onal requirements. Although the detailed requirements of the regulatory guide have not been incorporated as project criteria, review of the design criteria used for CGS i ndicates that the intent of this regulatory guide is met.
Specific Evaluation Reference
- See Sections 3.9.3.4 and 5.4.14.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 eval uation 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 detail s and documentation of data and studies required by the guide to suppor t such testing is not applicable.
Specific Evaluation Reference
- Not applicable.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 st ructures as defined in this regulatory guide do not exist.
General Compliance or Alternate Approach Assessment
- Not applicable.
Specific Evaluation Reference
- Not applicable.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 applicat ions docketed after December 1, 1977. However, CGS complies with the intent of th e guidance set forth in this re gulatory guide by an alternate approach.
General Compliance or Alternate Approach Assessment
- Safety-related battery installation desi gn 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 stri ct 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 di rectly applicable to CGS, Energy Northwest's maintenance pro cedures 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 T echnical 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 guidan ce 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 su rveillance. The standby diesel fuel oil storage tanks are protected with cathodic protection by anodes wh ich are located in th e near vicinity, but there are no pigtails connected to the fuel oil syst em 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 o il 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 ta nk 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 devi ates from the ANSI N195-1976 standard because of the pe rmanent interconnection between the standby power system and the auxiliary boiler system. The aux iliary boiler storage tank and its connective piping are not Safety Class
- 3. The auxiliary boiler st orage tank and its connecting auxiliary boiler system are not in a vital area, although AN SI 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 te sting activities to ensure that the minimum required volume is maintained.
Specific Evaluation Reference
- See Section 9.5.4.4.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 ga seous radioactive wast e systems were being designed and fabricated as ASME Section II I, 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 implem entation of this re gulatory guide these systems, and comp onents which follow the guidance found in the regulatory guide are indicated as Quality Clas s II+ and Code Group D+.
Specific Evaluation
Reference:
See Sections 3.2.4 and 3.2.6.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-202 Regulatory Guide 1.144, Re vision 1, September 1980 Auditing of Quality Assurance Pr ograms 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 require ment for the audited organization to investigate any adverse a udit finding to determine and schedule appropriate corrective action, Bech tel's auditing organization may determine the investigatory action and co rrective action including acti on 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 appropriat e 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 accepta ble methodology to determin ing site-specific relative concentrations for assessing the potential offs ite radiological conse quences 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 th e 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
.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-204 Regulatory Guide 1.146, Revision 0, August 1980 Qualification of Quality A ssurance Program Audit Personne l 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 qua lification comply with AN SI Standard N45.2.23.
II Operational Phase Compliance is discussed in the OQAPD.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 Guid e 1.147. The approval letter required that Energy Northwest document application of the code case in the FSAR.
The code case was first used fo r CGS in 1988 for deferral of hydr ostatic 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 Proj ect No. 2 (WNP-2)," dated August 8, 1985.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 desc ribe a method acceptabl e 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 spec ified 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. Sorens en, "Supplemental Safety Evaluation (SSE) of the Washington Public Power Supply System Nuclear Project No. 2 (WNP-2) Station Blackout Analysis (TAC M68626)," date d June 26, 1992.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 1.8-207 Regulatory Guide 1.160, Revision 1, January 1995 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.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 en sured by the implementation of a Control Room Envelope Habitability (CREH) Program and implementing procedures at CGS.
Specific Evaluation
References:
Not applicable.
COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT 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 en sured by the implementation of a Control Room Envelope Habitability (CREH) Program and implementing procedures at CGS.
Specific Evaluation
References:
Not applicable.