ML20249B408

Rev 6 to Trojan Nuclear Plant Defueled Sar

ML20249B408
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	06/17/1998
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20249B407	List: ML20249B406 ML20249B408
References
NUDOCS 9806230052
Download: ML20249B408 (80)

Text

1 TABLE OF CONTENTS DEFUELED SAFETY ANALYSIS REPORT j l

CHAPTER 1.0 I INTRODUCTION AND

SUMMARY

l Section Title i

Page l

1.0 INTRODUCTION

AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . 1.0-1 1.1 I nt rod uc tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1 1 l

1.2 General Plant Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 l 1.2.1 Design Criteria .................................. 1.2-1 1.2.2 Fuel Handling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-2 1.2.3 Radioactive Waste Treatment Systems . ................. 1.2-2 1.3 Identification of Agents and Contractors .................. 1.3-1 1.3.1 Design of the Facility ...... .. ............. ..... 1.3-1 l 1.4 Exclusions from and Exemotions to Certain Parts of Title 10 of the Code of Federal Regulations (10 CFR) . . . . . . . . . 1.4 1 1.4.1 Exclusions from Certain Parts of 10 CFR . . . . . . . . . . . . . . . . . . 1.4-1 1.4.1.1 10 CFR 26, Fitness for Duty Program . . . . . . . . . . . . . . . . . . 1.4-1 1.4.1.2 10 CFR 50.44, Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors . . . . . . . . . . . 1.4-1 1.4.1.3 10 CFR 50.46, Acceptance Criteria for Emergency Core l Cooling Systems for Light-Water Nuclear Power Reactors . . . . 1.4-2 1.4.1.4 10 CFR 50.48, Fire Protection ...................... 1.4-2 1.4.1.5 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1,1979 . . 1.4-2 1.4.1.6 10 CFR 50.49, Environmental Qualification of Electrical  ;

Equipment Important to Safety for Nuclear Power Plants . . . . . 1.4-2 1.4.1.7 10 CFR 50.55a, In-Service Inspection Requirements . . . . . . . . . 1.4-3 1.4.1.8 10 CFR 50.60, Acceptance Criteria for Fracture Prevention Measures for Light-Water Nuclear Power Reactors ... .. 1.4-3 1.4.1.9 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events . . . . . . . 1.4-3 i Revision 6 9806230052 980617 PDR ADOCK 05000344 W PDR L___-_--______-________-___--

CHAPTER 1.0 l INTRODUCTION AND

SUMMARY

l Section Title Page ]

l 1.4.1.10 10 CFR 50, Appendix G, Fracture Toughness Requirements ... 1.4-3 1.4.1.11 10 CFR 50, Appendix H. Reactor Vessel Material Surveillance Program Requirements . . . . . . . . . . . . . . . . . . 1.4-4 1.4.1.12 10 CFR 50.62, Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants . . . . . . . . . . . . . . 1.4-4 1.4.1.13 10 CFR 50.63, Loss of All Alternating-Current Power . . . . . . . 1.4-4 1.4.1.14 10 CFR 50.71(e), Maintenance of Record, Making of Reports . . 1.4-5 1.4.1.15 10 CFR 70.24, Criticality Accident Requirements . . . . . . . . . . 1.4-5 1.4.2 Exemptions to 10 CFR Related to the Permanently Defueled Condition . . . . . . . . ............. 1.4-5 1.4.2.1 10 CFR 50.54(o) and Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors .. ............... ............ 1.4-5 1.4.2.2 10 CFR 50.54(y), Cenditions of Licenses . . . . . . . . . . . . . . . . 1.4-6 1.4.2.3 10 CFR 50.54(q) and Certain Sections of 10 CFR 50.47,

" Emergency Plans," . . . . . . ..................... 1.4-6 ,

1.5 Material Incorocrated by Reference ................... 1.5-1 O I

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Revision 6 ii u

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CHAPTER 2.0 J

SITE CHARACTERISTICS Section Title Page 2.0 SITE CHARACTERISTICS . . . . . . . . . . . . . . . . . . . . . . . . . . 2.0-1 2.1 Geography and Demngraphv . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-1 2.1.1 Site Location and Description . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-1 2.1.1.1 Specification of Location . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-1  ;

2.1.1.2 Site Area Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-2 2.1.1.3 Boundwies for Establishing Effluent Release Limits . . . . . . . . . 2.1-3 3

2.1.2 Exclusion Area Authority and Control . . . . . . . . . . . . . . . . . . . 2.1-3 2.1.2.1 Authority . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-3 2.1.2.2 Exclusion Area Activities Unrelated to Plant Operation . . . . . . . 2.1-4 2.1.2.3 Arrangements for Traffic Control . . . . . . . . . . . . . . . . . . . . . 2.1-5

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1 2.1.3 Population Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-6 l 2.1.3.1 Population Within 10 Miles . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-6

/ 2.1.3.2 Population Between 10 and 50 Miles . . . . . . . . . . . . . . . . . . . 2.1-8 2.1.3.3 Transient Population . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-8  ;

2.1.3.4 Low-Population Zone ............................ 2.1-9 2.1.3.5 Population Center . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-10 1

2.1.4 Uses of Adjacent Lands and Waters . . . . . . . . . . . . . . . . . . . . . 2.1-10 2.2 Nearby industrial. Transportation and and Military Facilities . . . . . 2.2-1 2.2.1 Locations and Routes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-1

'2.2.2 Descriptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-4

.2.2.2.1 Description of Products and Materials . . . . . . . . . . . . . . . . . . 2.2-4 2.2.2.2 Pipelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-5 2.2.2.3 Waterways ................................... 2.2-6 2.2.2.4 A irports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-6 2.2.3 Evaluation of Potential Accidents . . . . . . . . . . . . . . . . . . . . . . . 2.2-7 2.2.3.1 Explosions ................................... 2.2-7 2.2.3.2 Toxic Chemicals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-17 2.2.3.3 Fires ....................................... 2.2-19 p 2.3.3.4 2.2.3.5 Ship Collision with Intake Structure . . . . ..............

Oil or Corrosive Liquid Spills in River . . . . . . . . . . . . . . . . .

2.2-20 2.2-21 iii Revision 6 l

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CHAPTER 2.0 SITE CHARACTERISTICS l Section Title Page 2.2.3.6 Cooling Tower Collapse . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-22 2.3 Meteo rolo gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3-1 2.3.1 Regional Climatology .... ........................ 2.3-1 2.3.1.1 General Climate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3-1 2.3.1.2 Regional Meteorological Conditions for Design and Operation Bases ....................... 2.3-1 2.3.2 Local Meteorology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3-2 2.3.2.1 Normal and Extreme Values of Meteorological Parameters .... 2.3-2 2.3.2.2 PotentialInfluence of the Plant and its Facilities On Local Meteorology . . . . . . . . . . . . . . . . 2.3-4 2.3.2.3 Local meteorological Conditions for Design and Operation Bases ....................... 2.3-5 2.3.3 Onsite Meteorological Measurements Program . . . . . . . . . . . . . . 2.3-6 g 2.3.3.1 Past Meteorological Facility Operations . . . . . . . . . . . . . . . . . 2.3-6 y 2,3.3.2 Measureme nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3-7 2.3.4 Diffusion Estimates . .............................. 2.3-8 2.4 Hydrologic Engineering . . . ............. ........... 2.4-1 2.4.1 Hydrologic Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-2 2.4.1.1 Site and Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-2 2.4.1.2 Hydrosphere . . ........ ..................... 2.4-2 2.4.2 Floods ...... ...... ........................ 2.4-3 2.4.2.1 Flood History .............................. .. 2.4-3 2.4.2.2 Flood Design Considerations . . . . . . . . . . . . . . . . . . . . . . . . 2.4-4 2.4.2.3 Effects of Local Intense Precipitation ..... ........... 2.4-5 2.4.3 Probable Maximum Flood of Streams and Rivers . . . . . . . . . . . . . 2.4-5 2.4.3.1 Probable Maximum Precipitation . . . . . . . . . . . . . . . . . . . . . 2.4-5 2.4.3.2 Precipitation Losses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-8 2.4.3.3 Runoff Model ............... . .............. 2.4-9 2.4.3.4 Probable Maximum Flood Flow . . . . . . . . . . . . . . . . . . . . . . 2.4-12 Revision 6 iv

CHAPTER 2.0 O

a SITE CHARACTERISTICS Section Title .P.ags-2.4.3.5 Water Level Determinations . . . . . . . . . . . . . . . . . . . . . . . . 2.4-18 2.4.3.6 Coincident Wind Wave Activity . . . . . . . . . . . . . . . . . . . . . . 2.4-19 2.4.4 Potential Dam Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-20 2.4.4.1 Seismically Induced Dam Failure . . . . . . . . . . . . . . . . . . . . . 2.4-20 2.4.4.2 Volcanically Induced Dam Failure . . . . . . . . . . . . . . . . . . . . 2.4-22 2.4.4.3 Spirit Lake Blockage Failure . . . . . . . . . . . . . . . . . . . . . . . . 2.4-25 2.4.5 Probable Maximum Surge Flooding . . . . . . . . . . . . . . . . . . . . . 2.4-29 2.4.5.1 Surge Water Levels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-29 2.4.5.2 Resonance .................................... 2.4-30 l 2.4.6 Probable Maximum Tsunami Flooding . . . . . . . . . . . . . . . . . . . 2.4-31 2.4.7 Ice Effects . . ................................... 2.4-31 2.4.8 Cooling Water Canals and Reservoirs . . . . . . . . . . . . . . . . . . . . 2.4-32 2.4.9 Channel Diversions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-32 2.4.10 Flooding Protection Requirements . . . . . . . . . . . . . . . . . . . . . . 2.4-32 2.4.11 Low Water Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-33 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters . . . . . . . . . . . . . 2.4-33 2.4.13 G rou ndwater . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-34 2.4.13.1 Description and Onsite Use . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-34 2.4.13.2 Sources . . . . . . . . . . . . . . . . . .................... 2.4-35 2.4.13.3 Accident Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-36 2.5 Geology. Seismology and Geotechnical Engineering . . . . . . . . . . . 2.5-1 2.5.1 Basic Geologic and Seismic Information ............... .. 2.5-1 2.5.1.1 Regional Geology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-2 2.5.1.2 S ite Geology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-4

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i CHAPTER 2.0 l SITE CHARACTERISTICS l l

Section Title Page 2.5.2 Vibratory Ground Motion . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-8 2.5.2.1 Seismicity . . . . . . . . . . . . . . . .................... 2.5-8 2.5.2.2 Geologic Structures and Tectonic Activity . . . . . . . . . . . . . . . 2.5-8 2.5.2.3 Maximum Earthquake Potential . . . . . . . . . . . . . . . . . . . . . . 2.5-9 2.5.2.4 Seismic Margin Earthquake, Safe Shutdown Earthquake and Operating Basis Earthquake . . . . . . . . . . . . . . . . . . . . . 2.5-11 2.5.3 Surface Faulting . . . . . . . . . . . . . . ................... 2.5-13 2.5.3.1 Geologic Condition of the Site . . . . . . . . . . . . . . . . . . . . . . . 2.5-13 2.5.3.2 Investigation for Capable Faults . . . . . . . . . . . . . . . . . . . . . . 2.5-14 2.5.3.3 Description of Faults . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-16 2.5.4 Stability of Subsurface Materials and Foundations . . . . . . . . . . . . 2.5-28 2.5.4.1 Foundation Evaluation . .................. ..... 2.5-29 2.5.5 Stability of Slopes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-33 2.5.6 Volcanology . . . . . . . ... ........................ 2.5-33 2.5.6.1 Volcanoes in General Area ........................ 2.5-34 2.5.6.2 Possible Hazards of the Cascade Volcanoes ........... .. 2.5-38 2.5.6.3 Hazards from Future Volcanic Activities . . . . . . . . . . . . . . . . 2.5-39 2.5.6.4 Summary and Conclusions . . . . . . . .. . . .. ....... 2.5-47 2.6 References ..... .................... .......... 2.6-1 Revision 6 vi U______.

A CEAPTER 3.0 V FACILITY DESIGN Section Title Eagc_

3.0 FACILITY DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.0-1  ;

i 3.1 Summarv ...................................... 3.1-1  ;

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3.1.1 Conformance with NRC General Design Criteria . . . . . . . . . . . . . 3.1-1 3.1.2 Classification of Structures, Components and Systems 3.1-7

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3.1.3 Wind and Tornado Loadings . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-7 3.1.3.1 Wind Loadings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-8 3.1.3.2 Tornado Loadings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-8 i 3.1.4 Water 1.evel (Flood) Design . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-13 3.1.5 Missile Protection Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-14 3.1.5.1 Missile Selection and Description . . . . . . . . . . . . . . . . . . . . . 3.1-14 3.1.6 Seismic Desig n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-15 3.1.6.1 - Seismic Design Parameters . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-15

-3.1.6.2 Seismic System Analysis .......................... 3.1-19 3.2 Spent Fuel Storage ................................ 3.2-1 3.2.1 Control, Auxiliary, and Fuel Building Complex ............. 3.2-1 3.2.1.1 Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.1.2 Desig n Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-6 3.2.1.3 Applicable Codes, Standards, and Specifications . . . . . . . . . . . 3.2-8 3.2.1.4 . Loads and Load Combinations . . . . . . . . . . . . . . . . . . . . . . . 3.2-13  ;

3.2.1.5 Design and Analysis Procedures . . . . . . . . . . . . . . . . . . . . . . 3.2-19 3.2.1.6 Structural Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . 3.2-28 3.2.1.7 Materials, Quality Control and Special Construction Techniques . . . . . . . . .......................... 3.2-30 3.2.1.8 . Testing and Inservice Inspection Requirements . . . . . . . . . . . . 3.2-34 3.2.1.9 Foundations . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . 3.2-34 3.2.2 Spent Fuel Pool and Fuel Storage Racks .................. 3.2-35 3.2.2.1 Design Bases . .... ........... .............. 3.2-35 O

V vii Revision 6

CHAPTER 3.0 FACILITY DESIGN Section Title Page 3.2.2.2 Sy stem Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-37 3.2.2.3 Design Evaluation . . . . . ..... ................... 3.2-38 3.2.2.4 Tests and Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-40 3.2.2.5 Instrumentation Application ........................ 3.2-40 3.2.2.6 SFP Structure Re-evaluation for Beyond Design Basis Seismic Motions . . . . . . . . . . . . . . . . . . . . . . 3.2-40 3.3 Auxiliary Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.1 Fuel Handling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.1.1 Design Bases . . . . . . . . . . . . ..................... 3.3-1 3.3.1.2 System Description . . . . . .................. .... 3.3-2 3.3.1.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-4 3.3.2 SFP Cooling and Demineralized System . . ................ 3.3-5 3.3.2.1 Design Bases . ................................ 3.3-5 3.3.2.2 System Description . . . . . . . . . . . . . . . . . ............ 3.3-6 3.3.2.3 Design Evaluation . . . . . . . . . . . . . . . ............... 3.3-8 3.3.3 Component Cooling Water System ..................... 3.3-8 3.3.3.1 Design Bases . . . .......................... ... 3.3-9 3.3.3.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-9 3.3.3.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-11 3.3.4 Service Water System . . . . . . . . . ... . .............. 3.3-11 3.3.4.1 Design Bases . . . . . . . . ......................... 3.3-12 3.3.4.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-13 3.3.4.3 Design Evaluation . . . . . . . . . . . . . . . . .............. 3.3-14 3.3.5 Compressed Air System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-15 3.3.6 Boric Acid Batch Tank . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3.3-15 1 3.3.7 Makeup Water Treatment System . . . . . . . . . . . . . . . . . . . . . . . 3.3-16 3.3.8 Equipment and Floor Drain Systems . . . . . . . . . . . . . . . . . . . . . 3.3-16 Revision 6 viii

i CHAPTER 3.0 t FACILITY DESIGN Section Title Page 3.3.9 Plant Discharge and Dilution Structure ................... 3.3-17 l 3.3.10 Primary Sampling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-18 3.3.11 Fire Protection System and Program . . . . . . . . . . . . . . . . . . . . . 3.3-18 3.3.12 Control Room Habitability . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-19 3.3.13 Seismic Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-19 3.4 Electric Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.1 Offsite Power System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.1.1 Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.1.2 Analys is . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-2 3.4.2 Onsite Power Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-2 3.4.2.1 Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-2 3.4.2.2 A naly s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-4 3.5 Compliance with NRC Regulatory Guides ................. 3.5-1 3.6 References ..................................... 3.6 1 f

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ix Revision 6 I

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1 CHAPTER 4.0 i OPERATIONS Section Title Page 4.0 O PERATION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.0-1 4.1 Operation Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1-1 4.1.1 Criticality Prevention . . . . . . . . ......... ...... .... 4.1-1 4.1.2 Chemistry Control ................................ 4.1-2 4.1.3 Instrumentation ............................... .. 4.1-3 4.1.3.1 Seismic Monitoring Instrumentation . . . . . . . . . . . . . . . . . . . 4.1 3 4.1.4 Maintenance Activities ..... ....................... 4.1-4 4.1.5 Administrative Control of Systems . . . . . . . . . . . . . . . . . . . . . . 4.1-5 4.2 Spent Fuel Handling . . . . . . . . ........ ............ 4.2-1 4.2.1 Spent Fuel Receipt, Handling, and Transfer ............... 4.2-1 4.2.1.1 Functional Description . . . . . . . . . . . . . . . . . . . ........ 4.2-1 #

4.2.1.2 ' Safety Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2-2 4.2.2 Spent Fuel Storage ................................ 4.2-3 4.3 Spent Fuel Cooling and Support Systems . . . . . . . . . . . . . . . . . . 4.3-1 4.3.1 Spent Fuel Pool Cooling ...... . ................... 4.3-1 4.3.1.1 Off-Normal Operation of the Spent Fuel Cooling System . . . . . . 4.3-2 4.3.1.2 Loss of Spent Fuel Pool Level . . . . . . . . . . . . . . . . . . . . . . . 4.3-2 4.3.1.3 Loss of Spent Fuel Pool Cooling . . . . . . . . . . . . . . . . . . . . . 4.3-3 4.3.1.4 High Spent Fuel Pool Level .. ..................... 4.3-4 4.3.1.5 Safety Criteria and Assurance ...................... 4.3-4 4.3.2 Electrical Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.3-5 4.3.3 Support Systems . . . . . . . . . . . . . . . . . . . . . . . . . ........ 4.3-5 l 4.4 Control Room Area . . . . . . . . . . . . . . . . ............... 4.4 1 j 4.5 References ......... ...... .... . .... .. .... 4.5-1 l

l Revision 6 x

CHAPTER 5.0

/7 RADIATION PROTECTION V Section Title Page

5. 0 RADIATION PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1 5.1 Source Terms ................................... 5.1-1 5.2 Offgns Treatment and ventilation . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.1 Vent Collection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.1.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.1.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-2 5.2.1.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-2 5.2.2 Containment Ventilation System . . . . . . . . . . . . . . . . . . . . . . . . 5.2-3 5.2.2.1 Des ign Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-3 l 5.2.2.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-3 5.2.2.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-5 5.2.3 Fuel and Auxiliary Building Ventilation System . . . . . . . . . . . . . . 5.2-5 5.2.3.1 Design Bases . . . . . . , , . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-5

(,, 5.2.3.2 System Description . . . .......................... 5.2-5 5.2.3.3 D.esign Evaluation . . . . . . . . . . . . . . . . . . ............ 5.2-6 5.2.4 Radwaste Processing Building Ventilation System . . . . . . . . . . . . 5.2-7  ;

5.2.4.1 Des ign B ases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-7  !

l 5.2.4.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-7 5.2.4.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-8 5.3 Liquid Waste Treatment and Retention ................... 5.3-1 5.3.1 Clean Radic, active Waste System . . . . . . . . . . . . . . . . . . . . . . . 5.3-1 5.3.1.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-1 l 5.3.1.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-2 l

5.3.1.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-4 l 5.3.2 Dirty Radioactive Waste System ....................... 5.3-5 5.3.2.1 Desig n Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-5 5.3.2.2 System Description . . . . . . . ..................... 5.3-6 5.3.2.3 Design Evaluation . . . . . . . ...................... 5.3-8 bv xi Revision 6

TROJAN NUCLEAR PLANT g PGE-1012 i FIRE PROTECTION PLAN j LIST OF EFFECTIVE PAGES Amendment Pace No. No.

Title Page 19 i through xiv 19 1.0-1 14 1.1-1 16 1.2-1 and 1.2-2 16 1.3-1 14 1.3-2 and 1.3-3 19 1.4-1 17 2.0-1 16 2.11 17 2.1-2 through 2.1-8 16 2.2-1 and 2.2-2 16 2.3-1 through 2.3-4 17 2.4-1 and 2.4-2 17 3.0-1 16 3.1-1 16 3.1-2 and 3.1-3 19 3.1-4 17 3.2-1 through 3.2-3 17 3.3-1 through 3.3-8 17 3.4-1 17 3.5-1 16 3.5-2 and 3.5-3 17 3.6-1 through 3.6-2 17 3.7-1 17 3.8-1 through 3.8-10 17 3.8-11 18 l

3.8-12 17 3.9-1 17 1

Amendment 19 xii el

CHAPTER 5.0 RADIATION PROTECTION Section Title Page 5.6.3.3 Radioactive Materials Safety . . . . . . . . . . . ............ 5.6-9 5.7 Ensuring that Occupational Radiation Exoosures are a low as is Reasonably Achievable (ALARA) ..................... 5.7-1 5.7.1 Policy Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7-1 5.7.2 Design Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7-1 5.7.3 Operational Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7-1 5.8 Collective Dose Assessment .......................... 5.8-1 5.9 References ..................................... 5.9-1

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xiii Revision 6

CHAPTER 6.0 ACCIDENT ANALYSIS Section Title Page 6.0 Accide nt A nalysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-1 6.0.1 Fission Product Inventories . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-1 6.0.1.1 Activities in the Core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-2 6.0.1.2 Activities in the Fuel Pellet Cladding Gap ............... 6.0-2 6.0.2 Radiological Evaluation Model ........................ 6.0-4 6.0.2.1 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-5 6.0.2.2 Whole Body Dose . . . . .......................... 6.0-5 6.0.2.3 Thyroid Inhalation Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-7 6.0.2.4 Computer Code ............. .................. 6.0-8 6.! Radioactive Release from a Subsystem or Component . . . . . . . . . . 6.1-1 I

6.1.1 Radioactive Gas Waste System Leak or Failure . . . . . . . . . .... 6.1-1 6.1.1.2 Assumptions or Conditions . . . . . . . . . . . . . . . . . . . . . . . . . 6.1-2  ;

6.1.1.3 Dose Results . . . . . . . . . . . . . . . . . ................ 6.1-2

~

6.1.2 Postulated Radioactive Releases Due l to Liquid Tank Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1-3 6.1.2.1 Identification of Causes and Accident Description .......... 6.1-3 6.1.2.2 Dose Results . . ....... ... ............ .. ... 6.1-4 6.2 Fuel Handling Accident . .. ........................ 6.2-1 6.2.1 Assumptions or Conditions . . ... ................... 6.2-1 6.2.2 Do se Resul ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.2-3 6.3 Spent Fuel Pool Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3-1 6.3.1 Loss of Spent Fuel Decay Heat Removal Capability . . . . . . .... 6.3-1 6.3.1.1 Potential Events Resulting in Loss of Spent Fuel Decay Heat Removal Capability ... .... ........ 6.3-1 6.3.2 Loss of Forced Spent Fuel Cooling Without Concurrent SFP Inventory Loss . . .... ... ... ... 6.3-4 O/!

Revision 6 xiv

CHAPTER 6.0 ACCIDENT ANALYSIS Section Title fg 6.3.3 Loss of Forced Spent Fuel Cooling with Concurrent M SFP Inventcry Loss . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3-5 6.4 References ..................................... 6.4-1

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O O xv Revision 6

l CHAPTER 7.0 CONDUCT OF OPERATIONS l i

Section Title _Eagt.

7.0 CONDUCT OF OPERATIONS . . . . . . . . . . . . . . . . . . . . . . . . 7.0-1 7.1 PGE Organi7ational Structure . . . . . . . . . . . . . . . . . . . . . . . . . 7.1-1 7.1.1 Management and Technical Support Organization ............ 7.1-1 7.1.2 Nuclear Division ................................. 7.1-1 7.1.2.1 Plant Organizations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.1-1 7.1.2.2 Supporting Organizations . . . . . . . . . . . . . . . . . . . . . . . .. 7.1-3 7.1.2.3 Review and Audit Organizations ..................... 7.1-5

' 7.2 Plant Procedures . . . . . . . . . . . . . . . . . . . .............. 7.2-1 7.2.1 Procedures Related to Nuclear Safety . . . . . . . . . . . . . . . . . . . . 7.2-1 7.2.2 Nuclear Division Manual . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-2 7.2.3 Pla nt Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-3 7.2.3.1 Administrative Orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-4 7.2.3.2 Operating Instructions . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-4 7.2.3.3 Periodic Tests . . . . . . . . . . . . . . . . . ............... 7.2-4 7.2.3.4 Fuel Handling Procedures . . . . . . . . . . . . . ............ 7.2-4 7.2.3.5 Maintenance Procedures . . . ......... ... ......... 7.2-5 7.2.3.6 Radiation Protection Procedures . . . . . . . . . . . . . . ....... 7.2-5 7.2.3.7 Chemistry Procedures . . . . . . . . . . . . . . . . . . . . ....... 7.2-5 7.2.3.8 Plant Safety Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-6 7.2.3.9 Temporary Procedures . . . . . . . . . .... ............ 7.2-6 7.3 Training ...................................... 7.3-1 7.3.1 Training for Certified Fuel Handlers . . . . . . . . . . . . . . . . . . .. 7.3-1 7.3.2 Training for Plant Staff . . . . . . . . . . . . .......... ..... 7.3-1 7.3.2.1 General Employee Training . . . . . . . . . . . . . . . . . . . . . . . . . 7.3-1 7.3.2.2 Fire Brigade Training ....... .................... 7.3-2 7.3.2.3 Other Training Programs . . . . . . . . . ........... ... 7.3-2 l 7.4 Emergency Plan . . .. ......... .... .... . .. .. 7.4-1

. > vision 6 xvi l

i CHAPTER 7.0 CONDUCT OF OPERATIONS Section Title Page 7.5 Decommissioning Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7,$.;

7.6 Trojan Nuclear Plant Security Plan /Tro_ian Nuclear Plant Security Force Training and Qualification Plan . . . . . . . . . . . . . . . . . . . . 7.6-1

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l l

xvii Revision 6

I l

CHAPTER 8.0 g TECHNICAL SPECIFICATIONS w Section Title Page 8.0 TECHNICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . 8.0 1 l

O Revision 6 xviii l

1 J

CHAPTER 9.0 QUALITY ASSURANCE Section Title Page 9.0 QUALITY ASSURANCE ............................ 9,0 1 O

I I

• i l

l xix Revision 6

LIST OF TABLES DEFUELED SAFETY ANALYSIS REPORT Number Title l

Trojan Cumulative Frequency Distribution of x /Q Values l 2.3-1 (September 1,1971 - August 31,1972) 2.3-2 Annual Average x /Q Values for Continuous Ground-Level Releases (Trojan Site Data September 1,1972 - August 31,1974) 2.3-3 Annual Average Deposition Values for Continuous Ground-Level Releases (Trojan Site Data September 1,1972 - August 31,1974)

I 2.3-4 Annual Average X /Q, Deposition and Plume Depletion Factor at Site Boundaries and Offsite Exposure Locations for Ground Level Releases (Trojan Site Data September 1,1972 - August 31,1974) 2.3-5 Maximum Annular Sector, Terrain Adjustment Factors Derived From NUSPUF With Building Wake Adjustment Divided by NUSOUT for Standard Population Distances of 0.5 Mile to 4.5 Miles 2.3-6 Terrain Adjustment Factors Derived form NUSPUF With Building Wake Adjustment Divided by NUSOUT for Special Instances 3.1-1 Design Wind Loads 3.1-2 Capability of Structures to Withstand a Tornado 3.1-3 Locations of Gas Storage Tanks 3.1-4 Analyzed SFP Load Drops and Missiles 3.1-5 Damping Values, Percent Critical Damping 3.1-6 Frequencies and Modal Effective Weights for Control, Auxiliary, and Fuel Building Complex 3.1-7 Comparison of Maximum Accelerations from Time-History and Response Spectrum Analyses of the Unmodified Control Building Complex 3.2-1 Live Loads 3.2-2 Calculated Results - Fuel Building 3.2-3 Calculated Results - Auxiliary Building 3.2-4 Calculated Results - Control Building 3.2-5 Significant Elevations in the Spent Fuel Pool 3.5-1 List of Pertinent Regulatory Guides for Defueled Status 4.3-1 Spent Fuel Pool Chemistry Specification and Sampling Schedule 5.5-1 Radiological Analysis Summary of Liquid Process Samples 1

Revision 6 xx

____A

i l

LIST OF TABLES l O DEFUET FD SAFETY ANALYSIS REPORT i

l Number Title i

6.0-1 Core and Gap Activities Based on Full Power Operation for 650 Days Full  !

l Power: 3565 MWt  !

6.0-2 Core Temperature Distribution 6.0-3 Breathing Rates 6.04 Physical Data for Isotopes 6.0-5 Accident Atmospheric Dilution Factors 6.2-1 Input Data for Calculation of Site Boundary Doses of a Fuel Handling  !

Accident I 6.2-2 Resultant Doses from Fuel Handling Accident and Comparison with 10 CFR 100 6.3-1 Spent Fuel Pool Performance During Loss of Forced Cooling 6.3-2 Dose Rates at Spent Fuel Pool at Reduced Water Levels O

I O xxi Revision 6

LIST OF FIGURES DEFUELED SAFETY ANALYSIS REPORT O j

. Number Title 1.1-1 Facility Location 3.1-1 Design Response Spectra Operating Basis Earthquake 3.1-2 Design Response Spectra Safe Shutdown Earthquake 3.1-3 Synthetic Acceleration Time History Normalized to 1.0 g 3.1-4 Comparison of the Acceleration Response Spectrum of the Synthetic Time History With the Trojan Design Spectrum for 1% Damping 3.1-5 Comparison of the Acceleration Response Spectrum of the Synthetic Time History With the Trojan Design Spectrum for 2% Damping 3.1-6 Comparison of the Acceleration Response Spectrum of the Synthetic Time History With the Trojan Design Spectrum for 5% Damping 3.1-7 Comparison of the Acceleration Response Spectrum of the Synthetic Time History With the Trojan Design Spectrum for 7% Damping 3.1-8 Finite Element Model - Control, Auxiliary, and Fuel Building 3.1-9 North - South Horizontal Response Spectra for Control Building Elevations 61 feet 0 inch and 65 feet 0 inch 3.1-10 North - South Horizontal Response Spectra for Auxiliary Building Elevation 61 feet 0 inch 3.1-11 North - South Horizontal Response Spectra for Fuel Building Elevation 61 feet 0 inch 3.1-12 North - South Horizontal Response Spectra for Control Building Elevation 117 feet 0 inch 3.1-13 North - South Horizontal Response Spectra for Auxiliary Building Elevation 117 feet 0 inch 3.1-14 East - West Horizontal Response Spectra for Control Building Elevations 61 feet 0 inch and 65 feet 0 inch 3.1-15 East - West Horizontal Response Spectra for Auxiliary Building Elevation 61 feet 0 inch 3.1-16 East - West Horizontal Response Spectra for Fuel Building Elevations 61 feet 0 inch 3.1-17 East - West Horizontal Response Spectra for Control Building Elevation 117 feet 0 inch 3.1-18 East - West Horizontal Response Spectra for Auxiliary Building Elevation l 117 feet 0 inch I 3.1-19 3-D Model for the Fuel Building Steel Superstmeture 3.2-1 Fuel Building - Plan Elevation 45' Revision 6 xxii O

l _- .. _________ _ ____________ _____________ ________-

LIST OF FIGURES DEFUEI FD SAFETY ANALYSIS REPORT l

l

• Number Title i

3.2-2 Spent Fuel Pool - Typical Details i 3.2-3 Auxiliary Building - Plan Elevation 45' 3.2-4 Auxiliary Building - Plan Elevation 45', Containment Abutment 3.2-5 Fuel Building - Plan Elevation 66' 3.2-6 Auxiliary Building - Plan Elevation 61' 3.2 Typical Section Through Auxiliary and Fuel Buildings 3.2-8 Typical Steel Framing -

3.2-9 Typical Steel Column Details 3.2-10 Floor Plans Showing Modifications 3.2-11 Control Building Floor Plan EL 45'-0" Showing Existing and Shear Walls 3.2-12 Control Building Floor Plan EL 61'-0" & 65'-0" Showing Existing and Shear Walls l 3.2-13 Control Building Floor Plan EL 77'-0" Showing Existing and Shear Walls 3.2-14 Control Building Floor Plan EL 93'-0" Showing Existing and Shear Walls 3.2-15 Equipment Location, Reactor and Auxiliary Buildings - Plan Below Ground Floor A 3.2-16 Equipment Location,' Reactor and Auxiliary Buildings - Plan Operating b Floor, Elevation 45' l

3.2-17 Equipment Location, Reactor and Auxiliary Buildings - Plan Elevation 61'  !

3.2-18 Equipment Location Reactor and Auxiliary Buildings - Plan Elevation 77' l 3.2-19 Equipment Location, Reactor and Auxiliary Buildings - Plan Operating Floor and Above L 3.2-20 Equipment Location, Reactor and Auxiliary Buildings - Section A-A 3.2 21 Equipment Location, Reactor and Auxiliary Buildings - Sections B-B, D-D, E-E, F-F 3.2-22 Equipment Location, Reactor and Auxiliary Buildings - Sections C-C and F-F 3.2-23 Containment Structure Typical Details 3.2-24 Containment Structure Typical Liner Plate Details 3.2-25 Containment Structure Base Slab Bottom Reinforcing 3.2-26 Containment Structure Base Slab Top Reinforcing 3.2-27 Containment Structure Wall Reinforcing 3.2-28 Containment Structure Dome 3.2-29 Containment Structure Typical Penetration Details 3.2-30' Containment Structure Prestressing Tendons at Equipment Hatch 3.3-1 Fuel Pool Cooling and Demineralized System l .

/ xxiii Revision 6

LIST OF FIGURES DEFUELED SAFETY ANALYSIS REPORT ,

Number Title I l

3.3-2 Component Cooling Water 3.3-3 Service Water System 4.2-1 Spent Fuel Storage Pool 4.2-2 Plant Arrangement Diagram of Fuel Cask Movement Envelope 5.2-1 Gaseous Radioactive Waste System 5.2-2 Containment Purge Supply System (CS-1) 5.2-3 Containment Purge Exhaust System (CS-2) 5.2-4 Fuel / Auxiliary Building Ventilation Supply System (AB-2) 5.2-5 Fuel / Auxiliary Building Ventilation Exhaust System (AB-3) 5.2-6 SFP Ventilation Exhaust System (AB-4) 5.2-7 Condensate Demineralized Building Ventilation Exhaust System 5.3-1 Clean Radioactive Waste System 5.3.2 Dirty Radioactive Waste System 5.3-3 Liquid Radwaste Demineralizers 5.4-1 Solid Radioactive Waste System 6.1-1 Waste Gas Storage Tank Rupture Whole Body (Beta plus Gamma) Dose 6.3-1 Decay Heat Generated from Stored Fuel 6.3-2 SFP Heatup Rate versus Time After Reactor Shutdown 6.3-3 Time for SFP to Boil Upon Loss of Forced Cooling 6.3-4 SFP Boil Off Rate Without Makeup versus Time After Reactor Shutdown 6.3-5 Makeup Rate to Maintain SFP Level During Boil Off versus Time After Reactor Shutdown 6.3-6 Boil Off Time to 10 Feet Above Fuel Versus Time After Reactor Shutdown l Revision 6 xxiv O

1

LIST OF EFFECTIVE PAGES DEFUFI Fn SAFETY ANALYSIS REPORT Section Effective Pages Revision Title Page N/A Rev.O Table of Contents i - xxiv Rev.6 List of Effective Pages xxv - xix Rev.6 1.0 1.0-1 Rev.0 1.1 1.1-1 Rev.0 1.2 1.2-1 Rev.0 1.2 1.2-2 Rev.4 1.3 1.3-1 Rev.0 1.4 1.4-1 through 1.4-4 Rev.4 1.4 1.4-5 Rev.6 1.4 1.4-6 Rev.4 1.5 1.5-1 Rev.5 Figure 1.1-1 N/A Rev.0 2.0 2.0-1 Rev.0 2.1 2.1-1 and 2.1-2 Rev.0 2.1 2.1-3 Rev.3 2.1 2.1-4 through 2.1-12 Rev.0 2.2 2.2-1 through 2.2-7 Rev.0 ,

2.2 2.2 8 Rev.4 l 2.2 2.2-9 Rev.O i 2.2 2.2-10 Rev.4 2.2 2.2-11 through 2.2-16 Rev.O I 2.2 2.2-17 Rev.3 i I

2.2 2.2-18 and 2.2-19 Rev.O i

2.2 2.2-20 Rev.6 2.2 2.2-21 and 2.2-22 Rev.0 2.3 2.3-1 through 2.312. Rev.0

' 2.4 2.4-1 Rev.3 2.4 2.4-2 and 2.4-3 Rev.0 2.4 2.4-4 Rev.3 2.4 2.4-5 through 2.4-22 Rev.0 2.4 2.4-23 Rev.6 2.4 2.4-24 through 2.4-31 Rev.0 2.4 2.4-32 Rev.4 2.4 2.4-33 through 2,4-35 Rev.0

\ xxy Revision 6

LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision 2.4 2,4-36 Rev.6 2.5 2.5-1 through 2.5-8 Rev.0 2.5 2.5-9 Rev.6 2.5 2.5-10 and 2.5-11 Rev.0 2.5 2.5-12 Rev.6 2.5 2.5-13 Rev.4 2.5 2.5-14 through 2.5-47 Rev.0 2.6 2.6-1 through 2.6-10 Rev.O Tables 2.3-1 through 2.3-6 N/A Rev.O Figure 2.3-1 N/A Rev.O Figures 2.4-1 and 2.4-2 N/A Rev.0 3.0 3.0-1 Rev.0 3.1 3.1-1 through 3.1-5 Rev.0 3.1 3.1-6 and 3.1-7 Rev.5 3.1 3.1-8 through 3.1-28 Rev.0 3.2 3.2-1 Rev.0 3.2 3.2-2 Rev.4 3.2 3.2-3 through 3.2-7 Rev.5 3.2 3.2-8 through 3.2-34 Rev.4 3.2 3.2-35 through 3.2-41 Rev.5 3.3 3.3-1 Rev.5 3.3 3.3-2 through 3.3-3 Rev.0 3.3 3.3-4 Rev.4 3.3 3.3-5 through 3.3-7 Rev.0 3.3 3.3-8 through 3.3-15 Rev.6 3.3 3.3-16 through 3.3-18 Rev.5 3.3 3.3-19 Rev.6 3.4 3.4-1 Rev.0 3.4 3.4-2 and 3.4-3 Rev.4 3.4 3.4-4 Rev.6 3.5 3.5-1 Rev.0 3.6 3.6-1 Rev.0 3.6 3.6-2 Rev.5 l 3.6 3.6-3 Rev.O Tables 3.1-1 and 3.1-2 N/A Rev.O Revision 6 xxvi

J LIST OF EFFECTIVE PAGES n DEFUELED SAFETY ANALYSIS REPORT 4

Section Effective Pages Revision Table 3.1-3 N/A Rev.6 Tables 3.1-4 through 3.1-7 N/A Rev.0 ,

Tables 3.2-1 through 3.2-3 N/A Rev.O I Tables 3.2-4 through 3.2-5 N/A Rev.1 Table 3.5-1 Sheet 1 Rev.3 l Table 3.5-1 Sheet 2 Rev.4 Table 3.5-1 Sheets 3 through 8 Rev.3 Table 3.5-1 Sheets 9 through 15 Rev.4 Table 3.5-1 Sheets 16 through 38 Rev.3 Figures 3.1-1 through 3.1-19 N/A Rev.O Figures 3.2-1 through 3.2-13 N/A Rev.O Figure 3.2-14 N/A Rev.6 Figures 3.2-15 through 3.2-22 N/A Rev.4 Figures 3.2-23 through 3.2-30 N/A Rev.O Figure 3.3-1 N/A Rev.5

Figures 3.3-2 and 3.3-3 N/A Rev.6 I I

V 4.0 4.0-1 Rev.4 4.0 4.0-2 Rev.3  ;

4.1 4.1-1 and 4.1-2 Rev.5  !

4.1 4.1-3 through 4.1-6 Rev.6 4.2 4.2-1 Rev.5 4.2 4.2-2 Rev.4 4.2 4.2-3 Rev.5 4.3 4.3-1 and 4.3-2 Rev.0 4.3 4.3-3 Rev.4 4.3 4.3-4 and 4.3-5 Rev.6 4.3 4.3-6 Rev.4 4.4 4.4-1 Rev.4 l

4.5 4.5-1 Rev.5 Tables 4.L-1 N/A Rev.5 Figure 4.2-1 N/A Rev.4 Figure 4.2-2 N/A Rev.4 5.0 5.0-1 Rev.0 5.1 5.1-1 Rev.0 5.2 5.2-1 through 5.2-3 Rev.O xxvii Revision 6 o

LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT O' '

Section Effective Pages Revision

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5.2 5.2-4 Rev.5 5.2 5.2-5 and 5.2-6 Rev.0 5.2 5.2-7 and 5.2-8 Rev.5 5.3 5.3-1 through 5.3-5 Rev.0 5.3 5.3-6 through 5.3-8 Rev.6 5.4 5.4-1 and 5.e-2 Rev.0 5.4 5.4-3 and 5.4-4 Rev.5 5.5 5.5-1 Rev.0 5.5 5.5-2 through 5.5-4 Rev.3 5.5 5.5-5 through 5.5-10 Rev.5 5.6 5.6-1 Rev.1 5.6 5.6-2 Rev.3 5.6 5.6-3 Rev.0 5.6 5.6-4 through 5.6-6 Rev.3 5.6 5.6-7 Rev.3 5.6 5.6-8 Rev.5 5.6 5.6-9 and 5.6-10 Rev.3 5.7 5.7-1 Rev.0 5.8 5.8-1 Rev.0 5.9 5.9-1 through 5.9-2 Rev.O Tables 5.5-1 N/A Rev.3 Figure 5.2-1 through 5.2-4 N/A Rev.5 Figures 5.2-5 and 5.2-6 N/A Rev.O Figure 5.2-7 N/A Rev.5 Figure 5.3-1 N/A Rev.5 Figure 5.3-2 N/A Rev.6 Figure 5.3-3 N/A Rev.O Figure 5.4-1 N/A Rev.0 6.0 6.0-1 through 6.0-8 Rev.0 6.1 6.1-1 Rev.5 6.1 6.1-2 through 6.1-4 Rev.0 6.2 6.2-1 through 6.2-3 Rev.0 6.3 6.3-1 Rev.5 6.3 6.3-2 through 6.3-4 Rev.0 6.3 6.3-5 Rev.4 6.3 6.3-6 Rev.0 4

Revision 6 xxviii

LIST OF EFFECTIVE PAGES f~) DEFUEI ED SAFETY ANALYSIS REPORT

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Section Effective Pages Revision )

6.3 6.3-7 Rev.4  ;

6.4 6.4-1 Rev.3 l 6.4 6.4-2 Rev.4 Tables 6.0-1 through 6.0-5 N/A Rev.O Tables 6.2-1 and 6.2-2 N/A Rev.O Tables 6.3-1 and 6.3-2 N/A Rev.O Figures 6.3-1 through 6.3-6 N/A Rev.0 7.0 7.0-1 R.ev.3 7.1 7.1-1 Rev.4 7.1 7.1-2 Rev.3 7.1 7.1-3 and 7.1-4 Rev.6 7.1 7.1-5 Rev.3 )

ough 7.2-6 0 7.3 7.3-1 and 7.3-2 Rev.3 7.3 7.3-3 Rev.0 7.4 7.4-1 Rev.0 7.5 Rev.4

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7.5-1 7.6 7.6-1 Rev.3 8.0 8.0-1 Rev.0 9.0 9.0-1 Rev.O i

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xxix Revision 6

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I 1.4.1.14 10 CFR 50.71(et Maintenance of Record. Making of Renorts v

This regulation applies to operating nuclear power plants. TNP has been defueled and is not authorized under the POL to operate as a nuclear power plant again. Thus, this regulation is not applicable to TNP. The Final Decommissioning Rule, which became effective August 28,1996, modified this rule to make it applicable to plants that are permanently shut j down, and altered the rule to recognize the limited safety importance of Safety Analysis Updates for shutdcwn plants. The modified final rule is, therefore, now applicable to Trojan.

1.4.1.;l,10 CFR 70.24. Criticality Accident Requirements l

By letter dated March 24,1993 from S. H. Weiss (NRC) to J. E. Cross (PGE), the staff has l 1

notified Trojan that with the absence of new fuel, the design of the storage racks, and the  !

procedural controls over fuel handling activities with Certified Fuel Handler training, adequate assurance against the occurrence of an accidental criticality exists that this regulation no longer applies to Trojan in its permanently shutdown and defueled condition.

1.4.2 EXEMPTIONS TO 10 CFR RELATED TO THE PERMANENTLY DEFUELED CONDITION There are also certain 10 CFR regulations to which the NRC has granted specific exemptions to PGE for the operation of the Trojan Facility.

1.4.2.1 10 CFR 50.54(o) and Appendix J. Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors By letter dated April 12,1993 from M. T. Masnik (NRC) to J. E. Cross (PGE) the NRC will permit PGE to cease all testing required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J.

O b 1.45 Revision 6 I

1.4.2.2 10 CFR 50.54(y). Conditions of Licenses By letter dated June 23,1993 from M. T. Masnik (NRC) to J. E. Cross (PGE) the NRC will permit PGE to depart from a license condition or technical specification in an emergency as described in 10 CFR 50.54(y) provided that the emergency action is approved, as a minimum, by a Certified Fuel Handler prior to taking the action.

l 1.4.2.3 10 CFR 50.54(a) and Certain Sections of 10 CFR 50.47. " Emergency Plans." and l 10 CFR 50. Appendix E. " Content of Emergency Plans" l

l By letter dated September 30,1993, from M. T. Masnik (NRC) to J. E. Cross (PGE), the l NRC authorized the Trojan facility to discontinue offsite emergency preparedness activities l and reduce the scope of onsite planning.

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l Revision 4 1.4-6 j

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l

In addition, a survey of major chemical shippers (72* using the Columbia River indicated that the principal toxic chemical shipped by river is anhydcous ammonia, with a frequency of 10-15 shipments per year".

' Accident analyses were conducted for those chemicals mentioned above for the operational life of the Trojan facility but with permanent defueling these analyses are no longer applicable.

The DBA for defueled operations either result in doses that are less than 10 CFR 100 limits (radioactive release from a subsystem or component and fuel handling accident) or do not require operator action for at least 4 days (loss of forced spent fuel cooling).

2.2.3.3 Fires Industries and oil storage facilities in the Trojan area are separated from the site, either by ,

considerable distance or by the Columbia River. Thus, there is no potential hazard to the site as a result of fires at these facilities.

O V

The site is protected from brush or forest fires on two sides by water, the Columbia River to the east and the Recreation Lake, Reflecting Lake and Chistling Swan area to the west.

Buildings on the site are afforded fire protection by the open areas and parking lots surrounding them.

l A fire caused by a rupture of the natural gas main west of the Plant would be separated from 1 the Plant buildings by a considerable distance and by the lake areas on the site. The Plant's Fire Protection System extends out to the Visitors Information Center and portable pumps could be used, operating from the lake areas of the site, in the event a fire were to spread down the hillside from the line.

O 2.2-19

2.2.3.4 Shio Collision with Intake Structure l

l The cooling water intake stmeture is a reinforced concrete structure that houses the pumps for the Service Water System (SWS). The structure is oriented perpendicular to the river flow on the west bank of the Columbia River, approximately 1400 feet west of the midstream ship channel center line. The entrance to the structure, which is set into the bank, is provided by the intake channel, an earthen cut approximately 110-feet long by 45-feet wide with side slopes of 3:1 extending to Elevation foot MSL. The river bottom in the vicinity of the intake structure and channel extends from elevation of about 3-foot MSL, adjacent to the front of the structure to an elevation of 0-foot MSL in about 70 feet. From there it drops off at a slope of approximately 1-foot vertical drop per 5-foot horizontal run to an elevation of -40 to foot MSL, approximately 200 feet from the mouth of the intake channel.

l The draft of ships using the river can be divided into four main categories: deep draft vessels with drafts of 30 to 40 feet, medium draft vessels with drafts of 20 to 30 feet, including unloaded deep draft vessels, shallow draft vessels with drafts less than 20 feet, consisting mainly of barges and pleasure craft with drafts of 1 to 10 feet.

The possibility of a ship collision with the intake structure depends on the size and draft of the ship involved, the stage of the river, the flow velocity of the river and whether the ship is drifting or powered. At normal river flows (stage Elevation 1-foot to 10.5-foot MSL), a vessel drifting without power would either be swept past the structure by the river current (with a velocity of 1 to 3 fps) or be grounded by the river bottom upstream of the intake structure. In order to impact with the structure, a vessel would have to be diverted from the midstream ship channel and aimed at the structure while under power with just the right timing to enter the relatively narrow intake channel while crossing a current of 1 to 5 fps. The effects of a channel itself must also be taken into account. A deep draft vessel (average draft 35 feet) would have to be narrower than the channel, and the river stage would have to exceed 22-foot MSL for the vessel to collide Revision 6 2.2-20 ]

l Swift Dam, near River Mile 49, is a rock-fill and earth-fill dam with a total storage of 755,000 acre-feet. The drainage area above Swift Dam is 481 sq miles.

Yale Dam, near River Mile 35, is a rock-fill dam with a total storage of 402,000 acre-feet and a drainage area of 5% sq miles.

Merwin Dam, near River Mile 21, is a concrete arch dam with a total storage of 405,000 acre- 1 feet and a drainage area of 730 sq miles. l All of the above dams are owned by Pacific Power & Light Company and are used for power generation.

Inspection reports made pursuant to the Federal Power Commission's Regulations under the Federal Power Act for dam projects along the Lewis River conclude that a PMF can be routed through the Lewis River without causing failure of any of the dams on that stream.

2.4.4.2.1 Dam Failure Permutations - Lewis River Flood i l

)

The proposed artificial flood is considered to be caused by a catastrophic event such as a mud

- flow or an extremely heavy fall of volcanic ash resulting from an eruption of Mt. St. Helens, lying to the north of the Swift Dam. These events in turn cause a massive wave in Swift Reservoir. This wave, upon reaching Swift Dam, completely and instantaneously breaches the dam. The flood is then routed downstream failing each dam in succession. Yale and Merwin Dams were considered to fail instantaneously when overtopped by 50 feet of water.

'- 2.4-23 Revison 6 L_

L l ,

l 2.4.4.2.2 Unsteady Flow Analysis - Lewis River Flood The methods used in the routing of the Lewis River artificial flood are considered extremely conservative. The instantaneous failure of Swift and Yale Dams is considered incredible due to the dam structures. A rock-fill or earth-fill dam fails by a washout of material (erosion) caused by water overflowing the nonoverflow sections of the dam. The mode of failure is therefore a slow breach of sections of the dam with attendant slower flows. When a considerable portion of the upper section of the dam is removed over a large time interval, the backwater reaches of the reservoir are reduced with attendant increases in tail water level, the driving force for dam failure is removed.

The analysis of the breach of the concrete arch section of Merwin Dam follows much the same methods. Failure of Merwin Dam would not be complete and instantaneous. An embankment north of the concrete section and the spillway gate sections would fail under lesser amounts of over-topping. The tail water elevation would then increase, reducing the stresses on the concrete-arch section of the dam.

Further conservatism is added by the analysis of the food wave as it enters the flood plain at the confluence of the Lewis River and the Columbia River. Conservative methods were used in detennining backflow up the Columbia River and storage in the upstream reaches of the Columbia River.

Base flows in the rivers were assumed as 10,000 cfs for the I.ewis River (twice the annual average and 340,000 cfs for the Columbia River).

2.4.4.2.3 Water Levels at Plant Site - Lewis River Flood For the Lewis River artificial flood, the peak water level at the site corresponding to the peak flow (3,300.000 cfs) will be from 39 to 41 feet MSL. Attendant wave mn up will be 1.7 feet.

2.4-24 4

I in the slough is between 14 feet anu 17 feet. The Goble Volcanics have poor aquifer characteristics.

1 The sicugh is filled with fine grained alluvium of Quaternary age. Previous seismic surveys done in the area indicate the bedrock to be up to 340-foot depth. None of the borings drilled in the deep section of the slough penetrated the full thickness of the alluvium. One of the deep borings - DH terminated about 9 feet in a gravel bed a 269-foot depth, which is overlain by 154 feet of fine sand,  ;

beneath a 114-foot-thick layer of organic silt with a trace of clay. A layer of volcanic ash occurs near 70-foot depth within the silt la'j er.

The entire alluvial section appears to be hydraulically connected to the river on both ends of the slough. Water levels in the two domestic water wells located near the south end of the alluvial channel respond to tidal fluctuations. A gravel bed below the fine sand is the aquifer for the two wells which supply the Trojan site. These wells are 8 inches in diameter and were drilled through the 158-foot-thick impermeable gray silt bed and 117 feet of fine sand into about 28 feet of gravel at the bottom of the alluvial channel. The gravel bed appears to be bimodal or strongly gap graded. It is composed of medium-to-coarse gravel with a very fine sand matrix. This fine sand controls the aquifer characteristics. These wells are each capable of producing approximately 250 gpm of high-quality domestic water.

l 2.4.13.2 Sources A survey of existing wells and natural springs was made in the area between Goble and the south edge of Rainier to determine the extent to which groundwater is presently utilized, and to determine the elevation of the permanent water table in the area of the site. The survey showed that bedrock supplies most of the groundwater to existing wells in the area. Approximate water levels in the existing wells were determined, and 2.4-35

l piezometers were installed in six of the drill holes at the site to indicate water levels in the alluvium and in the bedrock.

Static water levels in wells, and the elevation of U.S. Highway 30, show that water levels in the ridge are considerably higher than are the water levels in the alluvium, even during periods of very high flow in the river. Consequently, it is apparent that the water table in the alluvium does not feed the water table in the ridge. Thus, the precise local direction of movement of the water is not as important as is the fact that the water in the rock and in the alluvium moves toward the Columbia River and not toward and existing offsite wells or springs. The hydraulic gradient of the water table precludes contamination of the portion of the bedrock which now supplies groundwater to offsite wells or springs. It is therefore concluded that there is virtually no possibility of contamination of existing or future offsite groundwater supplies by accidental release of radioactive materials onto the alluvium or rock at the site.

2.4.13.3 Accident Effects Four permeability tests of the alluvial material in Drill Holes 9 and 10 showed permeability e

ranging from 10 feet to 20,000 feet per year. If accidental discharge of contaminated water onto the alluvium should occur, the water would move through the upper portion of the alluvium and toward the Columbia River at a rate of approximately 15 feet per year. If accidental discharge of contaminated water onto the foundation rock would occur the water would also move toward the Columbia River. If it moved through the fractures and pores in l the rock, it would move at a much slower rate than the rate in the alluvium.

1 Revision 6 2.4-36

foundation for the rritical structures. Unconfined compressive strengths of the 41 samples of tuffs, tuff breccias, and flow breccias (200 tons /sq. foot) with an average of 1225 psi (88 tons /sq. foot).

The specific gravity of the tuffs range from about 1.84 to 2.33 with about 2.10 the average.

Absorption ranges from 5 to 17 percent with an average of around 10 percent. Seismic Category I stmetures and Seismic Category II stmetures housing Seismic Category I equipment are generally

' founded on rock. In a few isolated instances, ponions of layers of compacted granular backfill over rock exist. Studies for liquefactive, thixotropy, or differential consolidation of soils were not required for these conditions.

Geophysical surveys showed compression wave velocities to be 8200 to 10,600 fps, which indicates adequate foundation conditions. Shear wave velocities of foundation rock for the reactor ranged from 4500 to 5000 fps. A value of 1.9 X 10' psi was used for the dynamic modulus of elasticity, based on the geophysical measurements. Values for static modulus of elasticity were obtained by numerous laboratory tests on representative core samples. A conservative value of 0.8 X 10' was used for the design. No uphole velocity measurements were made.

2.5.2.3 Maximinm Fnrthonske Potential i The largest historically recorded shock center within 50 miles of the site occurred on November 5, 1%2. It had an epicentral intensity of VII about 35 miles south of the site near Vancouver, Washington. At Longview, Washington, about 8 miles nonhwest of the site, and at Rainier, j Oregon, about 4 miles northwest of the site, the intensity was reponed as VI: however, the damage was confined to cracked plaster.

On October 12,1877, an intensity VII canhquake was felt in Portland, Marshfield (now called l l

l- Clackamas), and Cascades (now called Cascade Locks), Oregon. The location of the epicenter of this shock is uncenain. It is plotted 45 miles from the site in the southern part of Ponland, near Clackamas. The epicenter may have been farther east c

2.5 9 Revision 6 l

toward Cascade Locks, Oregon. The original reference to this canhquake is by Rxkwood*. j Unfortunately, he does not state exactly what his source is. His account follows:

" October 12,1877. Quite severe shocks were felt in Oregon occurring in Portland at 1:53 p.m. (Two shocks being noticed; at Marshfield, Clackamas Co., at I

1:45 p.m.; and at Cascades at 9:00 a.m.) The vibrations were in each case from north to sc di and were sufficiently violent to overthrow chimneys."

A second reference by Holden

• assigned an intensity VIII on the Rossi-Forel (R-F) scale to this earthquake. Townley and Allen

• also showed intensity VIII(R-F scale). Rasmussen*

and Berg

• both used intensity VIII but changed to the Modified Mercalli (MM) scale. The U.S. Coast and Geodetic Survey

• shows intensity VII. It is generally accepted that VII (MM) is the equivalent of VIII (R-F).

In correspondence with Bechtel, Mr. Don Tocher, Director of the Earthquake Mechanism L.aboratory of the U.S. Coast and Geodetic Su'/ey in San Francisco, stated that he believes an intensity VII (MM) to be correct for the October 12,1877, earthquake, ar$ states that the intensity VIII (MM) referred to by Berg

• and Rasmussen* is probably a res2lt of carelessness in designated scales when changing from the Rossi-Forel scale to the Modified Vercalli scale.

Seven earthquakes of maximum intensity VI were centered within 50 miles of the site. On February 3,1892, a shock of intensity VI o;' curred at Portland, and strong vibrations were felt at Astoria, Salem, and Lake Harney,235 miles southeast of Portland. On December 29, 1941, a shock of intensity VI was felt near Portland, Oregon, about 38 miles south of the site.

Another shock centered near Portland on December 15,1953, was felt with intensity IV at Kalama, Washington, about 2 miles southeast of the site. On September 15,1961, an intensity of VI was reported at Swift Dam on the Lewis River, which was designed for 0.10 g, but no damage was done to the dam. At Rainier 2.5-10 I

the shock on September 17,1961, was also of intensity VI but was not felt 30 miles south of the site. At Rainier, the intensity was reported about 30 miles southwest of the site. The intensity was reported as V in Lengview,7 miles north of the site, and III at Goble. Oregon, about 1 mile south of the site. It was not felt in Rainier.

The largest earthquakes within 150 miles of the site were two shocks of epicentral intensity VIII which occur =d on A,nril 13,1949, and April 29,1%5. The epicenters were in the Puget Sound Area, approximately 70 miles and 95 miles, respectively, northeast of the site. Heavy damage, deaths, and injuries were reported in the epicentral areas. The accelerograms indicate a maximum resultant horizontal acceleration cf about 0.10 g at Seattle, where the intensity was VIII. At Rainier, Oregon, which is about the same distance from the epicenter as is Seattle, the intensity was also given as VIII, which is the greatest intensity reported historically at Rainier from any earthquake. However, the intensity at Rainier was based on damage to only one building, and that building was founded on marshy ground. A study of the damage reported at Rainier indicates that a lower intensity might reasonably be assigned. At Goble, about 1 miles south of the site, the intensity due to the 1949 earthquake was only VI, apparently because Goble is founded on rock. The Trojan Plant is founded on rock belonging to the same unit as the rock that underlies Goble.

The April 29,1965, earthquake cause.1 lower intensities in the site area than the 1949 earthquake. The intensity at Kelso and Longview was VI; at Rainier it was V; and at Goble only IV.

2.5.2.4 Seismic Margin Earthquake, Safe Shutdown Earthquake and Ooeratine Basis Farthonske The Trojan Plant is located in an area that experiences moderate seismic activity. The records i show that the maximum intensity that has been reported at Rainier, Oregon,4 miles north of j the Site, is VII. Since this intensity occurred on overburden, it is probable 2.5-11 i

I that on rock at the Trojan site the intensity for this same shock was not over VII. Intensity VII correlates with a horizontal acceleration of 0.12 g according to Hershberger"*. This historical l

data formed the bases for assigning the Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (OBE) used for the safety analysis while Trojan Plant was in operation.

l The SSE was determined such that any probable earthquake experienecJ at the site would not exceed the intensity selected. An intensity of VIII was selected since it was probable that an earthquake of that magnitude had never been experienced at the site. An intensity VIII is equivalent to an acceleration of 0.25 g. The design cf the Trojan Plant was based on an SSE of 0.25 g.

The OBE was determined by historical occurrence of eanhquakes. As discussed above the maximum earthquake experienced st the Trojan Plant site was assumed to be an intensity VII.

Intensity VII correlates to 0.12 g, however, a more conservative value of 0.15 g was assigned for the OBE.

There have been significant changes in the perception of earthquake hazards in the Pacific Northwest since the time of the initial design and licensing of the Trojan Plant. It is now commonly believed among the geoscience community that large subduction zone earthquakes likely occurred along the Oregon-Washington-Vancouver Island coast (known as the Cascadia margin, or Cascadia subduction zone) within the recent past, and that the potential for such events to occur in the future should be considered in any evaluation of safety and reliability of critical facilities during earthquake loading.

In M87, in response to the emerging issue of potential subduction zone earthquakes, PGE initiated a program of close monitoring of earthquake hazard research, conducted along the Cascadia margin. The results of these studies, together with studies initiated by PGE, have l been used to characterize the maximum events that could be expected to occur in the region and the resulting free-field ground motions that may occur at the Revision 6 2.5-12

l i

(9) Valves and piping.

(10) Instrumentation.

The SFP cooling and demineralized system components are located in the Auxiliary Building and Fuel Building.

The SFP cooling and ciemineralizer system is a closed-loop system consisting of three subsystems:

)

the cooling system, the purification system, and the skimmer system. 1 The SFP cooling subsy; tem utilizes two SFP cooling pumps installed in parallel and two SFP l

cooling water heat exchangers installed in parallel with a common pipe between the pumps and the heat exchangers. The SFP cooling pumps draw suction from the SFP through a common strainer located above the spent fuel assemblies and discharge back to the SFP through the SFP cooling heat )

exchangers.

l The SFP purification subsystem utilizes the SFP purification pump to divert a small fraction of the total flow through the SFP purification filter and/or the SFP demineralized. Both the filter and )

demineralized have sufficient capacity to recirculate the entire SFP water inventory at least once every two days. This purification loop is adequate for removing fission products and other coritaminants which may be introduced by a leaking fuel assembly. Portable vacuum filtration units may also be used to support special cleanup evolutions.

The skimmer subsyst:m consists of the SFP skimmer pump and two 1-1/2-inch dimeter standpipe and valve assemblies with quick disconnects to which a floating skimmer can be connected by a flexible hose. The SFP skimmer pump can draw suction from the top water surfaces of the SFP and discharge through the skimmer filter or empty filter housing back to de SR.

T 3.3-7 L_ __--

3.3.2.3 Design Evaloah0D t

l A single SFP cooling pump and a single SFP cooling water heat exchanger can maintain SFP l temperatures s140 F.

All lines entering the SFP which could siphon the pool to Elevation 76 feet 7 inches (approximately 10 feet above fuel elements) or below are equipped with siphon breakers at Elevation 83 feet 11 inches.

The SFP cooling and demineralized system does not perform any safety functions. Loss of forced SFP cooling will cause the SFP water temperature to slowly rise. The longest time interval between SFP cooling system failure and its detection is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since the system is inspected once per shift.

The failure would be detected sooner if the SFP water temperature reaches the alarm setpoint. If forced cooling cannot be restored, then the SFP water temperature will continue to rise, increasing the evaporation rate and possibly resulting in boiling within the SFP. The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to the SFP and determined that sufficient time exists to effect repairs to the cooling system or to establish makeup flow prior to uncovering the spent fuel. Makeup water is available from a variety of sources as described in Section 4.3.1.

In the event of a design basis seismic event, the systems that were designed to Seismic Category 11 requirements may not be operable. In this case, sufficient time would be available to align a source of makeup water to maintain water level.

3.3.3 COMPONENT COOLING WATER SYSTEM The Component Cooling Water System (CCWS), shown in Figure 3.3-2, is a closed cycle system designed to provide a monitored intermediate barrier between components handling radioactively Revision 6 3.3-8

contaminated fluids and the primary heat sink (SWS). With the reactor permanently defueled, the CCWS provides the required cooling water for removal of heat from the SFP via the SFP cooling water heat exchangers. l With the reactor permanently defueled, the CCWS no longer performs any safety functions.

3.3.3.1 Design Bases Each loop of the CCWS was originally designed for design basis accident heat. loads and normal reactor operation heat loads, including the heat loads from both SFP cooling water heat exchangers (1.3 x 107 Bru/hr) m, The current heat load from the SFP cooling water heat exchangers is a small fraction of the original design basis CCWS heat load ensuring that the B-train CCWS loop can l provide the required heat removal capacity.

3.3.3.2 System Description O

V The CCWS is shown in Figure 3.3-2. The components included in the system are:

(1) One CCWS pump (P-210D). l (2) One CCWS heat exchanger (E-204B). l (3) Two CCWS surge tanks (T-204A and B).

(4) One chemical additive tank (T-209). j (5) Two CCWS makeup pumps (P-218A and B). l (6) Interconnecting piping, valves, and instrumentation.

l The CCWS is a closed-cycle cooling system with the fluid being continuously recirculated through l the system by the CCWS pump. No part of the system is normally open to the atmosphere. Heat is l removed from the system by the flow of service water through the tube side of the CCWS heat

{

exchanger. l 0 -

3.3-9 Revision 6

l P-210D provides the required circulation through the system and was sized for current heat loads.

l P-210D is supplied from 480-V MCC bus B-35. h l One CCWS heat exchanger provides the required heat transfer from the CCWS to the SWS. The service water, which has a greater tendency to foul heat exchanger surfaces, circulates through the l heat exchanger tubes to facilitate cleaning of the water surfaces. The CCWS heat exchanger is l 1

designed to provide the required heat removal capacity assuming a maximum service water temperature of 75*F.

l The primary function of the surge tanks is to provided a static pressure on the CCWS sufficient to maintain the fluid pressure throughout the system above the fluid vapor pressure. The tanks also provide the following functions: (a) a means for damping transient pressures, (b) a means of monitoring the volume of fluid in the system, (c) a means of providing for expansion and contraction of the fluid in the system,(d) a source of pressure relief to the system through the safety relief valves on the tanks to the dirty radioactive waste treatment system (DRWS).

l A cushion of nitrogen is used to provide pressure in the CCWS surge tanL. The nitrogen is normally supplied from the Plant nitrogen storage system. Self-contained pressure-regulating valves are installed in the nitrogen supply lines between the nitrogen sources and the surge tanks.

The pressure may also be adjusted manually using the solenoid valves installed in the nitrogen supply lines from the Plant nitrogen storage system. Each tank is equipped with two safety relief valves to prevent overpressurizing the system.

Provisions exist for adding corrosion inhibitor to the system.

Two identical CCWS makeup pumps are installed to fumish makeup water to the system. Normal makeup water is supplied from the demineralized wate. storage tank using the demineralized water transfer pumps.

Revision 6 3.3-10 O'

For the defueled condition, train independence and automatic isolation of selected loads are not I

required. A single CCWS pump and a single CCWS heat exchanger provide excess heat removal capacity for the current heat loads.

Piping in the portions of the CCWS providing SFP cooling is seamless carbon steel, fabricated and j installed in accordance with the requirements of ANSI B.31.1.0, Code for Power Piping.

3.3.3.3 Desian Evaluation The CCWS does not perform any safety functions. Loss of component cooling water to the SFP cooling water heat exchangers will cause the SFP water temperature to slowly rise. If the component cooling water cannot be restored to the heat exchangers, then the SFP water temperatum will continue to rise, increasing the evaporation rate and possibly resulting in boiling within the SFP. The only requirement to assure adequate cooling of the spent fuel is to maintain the water l level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to i

the SFP and determined that sufficient time exists to effect repairs to the cooling system or to establish makeup flow prior to uncovering the spent fuel. Makeup water is available from a variety of sources as described in Section 4.3.1.

3.3.4 SERVICE WATER SYSTEM i The SWS, shown in Figure 3.3-3, is designed to provide water from the Columbia River to cool equipment and to supply water to various systems and equipment. With the reactor permanently _

L defueled, the primary functions of the SWS are reduced to providing cooling water to the l component cooling water heat exchangers and selected room cooler units as well as provide makeup to the spent fuel pool.

O 3.3 11 Revision 6 1

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l l

_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ - . _ _ _ _ _ _ _ - . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___________-___-J

3.3.4.1 Desian Bases O

The SWS is designed to deliver the minimum required flows of water to equipment assuming a minimum water level of 1.5 feet below MSL in the Columbia River. With the reactor pennanently defueled, system design requirements are reduced substantially from the original design bases of the system.

Heat transfer equipment was selected based on a temperature of 75"F, which exceeds the highest recorded Columbia River water temperature.

The system design includes provisions for inhibiting long-term corrosion and organic fouling of the system water passages.

O Revision 6 3.3-12 O

3.3.4.2 System Description O

The SWS is shown in Figure 3.3-3. The components included in the system are:

. (1) Three SWS pumps (P-108A, B, and C).

(2) Two SWS strainers (F-101 A and B).

1

. (3) Two SWS booster pumps (P-148B and D). '

(4) Interconnecting piping, valves, and instrumentation.

(5) Two cooling water pumps (P-167A and B). l Water is supplied through the trash rack and traveling water screens at the intake structure. The water entering the system is periodically chlorinated for microbiological control.

One traveling water screen is installed in the flow path to each of the two independent flow paths to l the service water pumps at the intake structure. The screens are automatically cleaned by the screen l' wash system which consists of two vertical pumps taking suction on the downstream side of the I screens and discharging into high velocity spray nozzl.es which clean the debris from the screens as they travel past the nozzles. The screen wash system is automatically actuated by an adjustable timer or when there is a high differential level across the screen.

L l .Three identical service water pumps take suction from the river through the traveling water screens at the intake structure. Each pump was designed to provide 100 percent of the operating plant flow l

requirements. Pump P-108A is aligned to loop A and pump P-108B is aligned to loop B. Pump P-l 108C can be aligned to either loop by opening normally locked-closed valves. Power to P-108A is supplied from 4160-V bus Al and P-108B is supplied from 4160-V bus A2. Power to P-108C can 3.3-13 Revision 6 i

1 I

be supplied from either bus Al or A2 through a manually operated transfer switch. Lubricating water is supplied to the three service water pumps. River water can also be supplied to the SWS l using Cooling Water Pumps P-167A and P-167B. Either pump is able to supply,in excess, the l cooling water flow requirements of the SWS. P-167A can supply makeup flow to the SFP.

Two identical service water strainers are provided for straining the discharge from the service water l pumps and the cooling water pumps. One strainer is located in the supply header to each loop.

Two identical service water booster pumps, P-148B and D, are provided to boost the pressure in the water supply lines to components served by the system except the supplies to the CCWS heat exchangers, makeup water to the CCWS, bearing tube water and SWS strainer backwash. These are l supplied directly from the service water pumps. Either service water booster pump is capable of l providing in excess of 100 percent of the loop flow requirements.

A source of SFP makeup water is provided from the service water booster pumps.

For the defueled condition, train independence and automatic isolation of the Seismic Category II loads are not required. A single SW loop provides excess capacity for the current loads.

Discharged water is dechlorinated at the Plant discharge and dilution structure before being discharged to the Columbia River as discussed in Section 3.3.9.

3.3.4.3 Design Evaluation During normal operation, the only requirement for the SWS is to maintain the SFP temperature sl40 F.

Loss of the SWS will cause the SFP water temperature to slowly rise due to loss of heat removal capability from the SFP cooling water heat exchangers (due to loss of component cooling water O

Revision 6 ,.3-14

cooling). The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to the SFP and determined that sufficient time exists to effect repairs to the cooling system or to establish makeup flow from an alternate makeup source prior to uncovering the spent fuel.

3.3.5 COMPRESSED AIR SYSTEM The compressed air system provides the Plant compressed air requirements for pneumatic instruments and valves and for service air outlets located throughout the Plant which are used for operation of pneumatic tools. The system does not perform any safety functions.

! The system uses water-cooled aftercoolers and compressors. The air receivers are connected to a common compressed air header which connects to the air filter unit. The discharge of the air filter unit connects to the air-dryer unit inlet and the service air header. The instrument air header is connected to the air-dryer unit discharge. Each air header supplies branch lines which supply V .

l instmment air and service air to the required loads throughout the Plant. The instrument and service l

l air system provides air to the inflatable seals for the SFP gates and to the CCWS air-operated isolation valves (CV-3303, CV-3287, CV-3304, and CV-3288). Loss ofinstrument air to the j CCWS isolation valves would cause them to fail closed causing a loss of forced cooling to the SFP I l

! (due to loss of component cooling water to the SFP cooling water heat exchangers). The only I requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to the SFP and l ' determined that sufficient time exists to effect repairs to the cooling system or to establish makeup l flow prior to uncovering the spent fuel.

3.3.6 BORIC ACID BATCH TANK The boric acid batch tank will normally be used to supply borated water to the SFP. Procedural controls will be used for this process.

O 3.3-15 Revision 6

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3.3.7 MAKEUP WATER TREATMENT SYSTEM O

l The Makeup Water Treatment System provides demineralized water of the required quality to meet Plant needs. Makeup water is processed and then stored in the demineralized water storage tank where it is available to meet Plant needs. The DWST is a source of SFP makeup water. The water is transferred to the SFP using a demineralized water transfer pump. l 3.3.8 EOUIPMENT AND FLOOR DPAIN SYSTEMS l

The following equipment and floor drainage systems are provided for the Plant:

(1) Dirty Radioactive Waste Treatment System (DRWS) drains. I (2) Clean Radioactive Waste Treatment System (CRWS) drains.

(3) Oily waste system.

(4) Acid waste system.

(5) Sanitary waste system.

The equipment and floor drain systems do not perform any safety functions.

The DRWS and CRWS are designed to collect liquid waste from areas containing equipment that handles radioactive fluids. These systems are designed to control the spread and release of radioactive particulate by directing potentially radioactive fluids to the radioactive waste treatment systems. These systems are described in Section 5.3.

Revision 5 3.3-16 O

3.312 CONTROL ROOM HABITABILITY O .

To support habitability, the original control room design included radiation shielding, air filtering, l air conditioning and ventilation systems, fire protection, personnel protective equipment and first aid, and utility and sanitary facilities. As discussed in chapter 6, the only accident requiring operator action is a prolonged loss of SFP cooling. Since this postulated event does not require operator action for several days, short term actions initiated from the control room to restore SFP cooling or to establish SFP makeup water flow are not required to protect the health and safety of the public. Those systems originally provided to assure habitability during accidents are no longer required.

3.3.13 SEISMIC INSTRUMENTATION Seismic instmmentation for the facility consists of a multielement seismoscope and peak l acceleration recorders. The multielement seismoscope is mounted on an essentially rigid l foundation, which will provide iso significant amplification of ground motion. Peak acceleration l recorders are also installed at appropriate locations in the facility. The seismic instrumentation satisfies 10 CFR 100, Appendix A, which requires instrumentation so that the seismic response of features important to safety can be determined promptly to permit comparison of such response with that used as the design basis.

3.3-19 Revision 6

For defueled conditions, train separation is no longer required. The aoility to cross-tie buses provides flexibility for maintaining power to station loads.

i 3.4.2.1.3 480-V System l The 480-V system is designed to provide sufficient electric power for operation of Plant loads from load centers and MCC buses. The system consists ofload centers, MCCa, loads fed from these load

! centers and MCCs, interconnecting cables, and associated instrumentation and control circuits.

l For defueled conditions, train separation is no longer required. The e.bility to cross-tie buses j . provides flexibility for maintaining power to station loads.

~

3.4.2.1.4 120-V A-C System L The 120-V instrument a-c and the preferred instmment a-c systems are designed to provide reliable electric power for control and instrumentation.

i i

The 120-V instrument a-c system was designed to supply 208/120-V control and instrumentation I power for equipment not r: quired for the safe shutdown of the reactor. This system supplies power

- to the normally open service water isolation valves at the Seismic Category I (original design) l interface with the Seismic Category II portion of the system. In the defueled condition this l L automatic isolation is no longer needed.

l The 120-V preferred instrument a-c system consists of four 120-V bus sections which are supplied from'120-V instrument a-c buses. For defueled conditions, the system no longer performs any safety functions.

I

() 3.4-3 Revision 4

3.4.2.1.5 125-V D-C System O

The 125-V d-c system consists of two systems. Each system is comprised of one 125-V battery, two battery chargers, one 125-V d-c control center with two distribution panels and a split bus connected together by a nonautomatic circuit breaker, interconnecting cables, and associated instrumentation and control circuit. The system is arranged so that the battery or any one charger can independently supply the system bus. Power to each 125-V d-c system is normally supplied from one battery charger which rectifies 480-V a-c obtained from an associated 480-V a-c MCC. For defueled conditions, the system i no longer performs any safety functions.

3.4.2.2 Analysis A variety of electrical distribution system faults can cause the loss of forced cooling to the SFP. The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to the SFP and determined that sufficient time exists to effect repairs to the cooling system or to establish makeup flow prior to uncovering the spent fuel.

Based on the results of the analysis and the flexibility of the offsite and onsite electrical distribution system, sufficient time would b: available to perform one of the following options:

(1) Restore otTsite power and forced SFP cooling.

l (2) Provide attemate source of power to systems required for SFP forced cooling.

(3) Provide an a-c independent source of makeup water.

l l

Revision 6 3.4-4

TABLE 3.1-3 LOCATIONS OF GAS STORAGE TANKS l Equip. No. Service Location Dimensions psig T-118 Compressed Turbine Bidg. Elevation 3-ft dia x 165 l A,B,C air receivers 45 ft. outside EDG rooms 8-ft high max between column 51 -55 T-151 N2bulk gas Roofof Control Bldg. 24-in dia x 2450 A,B,C,D, tanks 21-ft long E, F T-159 Liquid CO2 Turbine Bldg. between 4-ft dia x 300 storage tank column 77 - 83, Elevation 10-filong 45 ft.

T-125 Hydro- Turbine Bldg. south end, Il-ft dia x 125 pneumatic Elevation 45 ft., between 19-ft long pressure column U and T accumulator i (air)

T-204 A, B CCWS surge Elevation 77 ft., column 7-ft dia x 150 I

( tank (N2) line E and 61, between 8-ft 3-in high Reactor and Auxilary Building O Revision 6

I ,

i I TABLE 3.1-4 ANALYZED SFP LOAD DROPS AND MISSILES O

i LOAD DROPS Max. Max. 3 D Drop Impact Impact i Item Wei ht Hei Energy Surface l No. Item Description (I n.) ght(i (in.-lbs) (in.)

1 Spent fuel assembly with 1,624 12 19,488 8.42 x 8.42 reactor control rod 2 Spent fuel assembly 356 172 61,200 9.0x9.0 handling tool 3 Spent fuel assembly with 1,976 12 23,700 8.42 x 8.42 handling tool 4 BPRA handling tool 800 148 118,400 18.0 x 18.0 5 Thimble plug removal 290 192 55,700 8.69 x 8.69 tool 6 Fuel assembly channel 350 254 88,900 8.42 x 8.42 spacing tool 7 Radiation specimen 600 12 7,200 8.42 x 8.42 transfer basket 8 New fuel tool 80 480 38,400 9.0 x 9.0 9 Electric chain hoist 90 530 47,700 8 diameter 10 Portable RCCA change 900 148 133,200 8.42 x 8.42 tool 11 Miscellaneous hand 500 300 150,000 36 x 48 tooling / portable vacuum system TORNADO MISSILES Vertical Velocity at item SFP Surface No. Item Description Dry Ib)

(Weight (mph) 1 4-in. x 4-in. x 12-ft long wood plank 108 140 2 3-in. diameter x 10-ft long steel pipe 76 52.5 l l I

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Corrective measures for out of specification chemistry will be taken upon discovery. The spent fuel purification subsystem is used to maintain satisfactory SFP purity and clarity. The i capability for boric acid addition is required to ensure SFP coolant can be maintained

>_.2000 ppm. Boric acid normally will be mixed in the boric acid batch tank and gravity-drained to the SFP.

4.1.3 INSTRUMENTATION The primary instrumentation associated with operation of the Plant is associated with the SFP cooling system. This instrumentation provides the operating staff with indication of SFP level, pump suction and discharge pressure, and temperatures throughout the system.

Alarms are provided to the control room for abnormal SFP level or high SFP temperature.

There are no automatic actions performed by SFP cooling instrumentation with system operation being manually controlled. A more complete description of the SFP cooling system is provided in Section 4.3.

4.1.3.1 Seismic Monitoring Instrumentation The seismic monitoring instrumentation, consisting of a multi-element seismoscope and peak l acceleration recorders, is subject to routine maintenance and testing to ensure it is available to l )

record data from seismic events. l l

SR-6341 Multi-element Seismoscope l SR-6340D Peak Acceleration Recorder - Control Building Mezzanine, Top of l Ladder Above Secondary Sample Storage Room l l SR-6340E Peak Acceleration Recorder - Fuel Building, 93-foot Elevation Hot l Shop, West, Behind the Wall l )

1 SR-6340F Peak Acceleration Recorder - CCW Heat Exchangers,45-Foot Elevation l Area 3 Base l 4.1-3 Revision 6 l

l l

l In the ' case where an instrument is removed from service for greater than 30 days, alternate l means shall be provided to record the data needed for comparison with the design bases of l l facility features important to safety, unless it is determined by Engineering that the remaining i instruments can provide sufficient data fcr analysis. Seismic monitoring instrumentation shall l be in service during spent fuel cask loading operations. While it remains installed, maintained

{ and tested, the time-history recording system provides an acceptable alternate means to record l the data needed from a seismic event.

l l ST-6336C Accelerograph - Fuel Building,93-Foot Elevation l ST-6336D Accelerograph - Cable Spreading Room l ST-6336E Accelerograph - Free Field l SR-6336A Seismic Recorder Unit l SR-6336B Seismic Recorder Unit l SR-6337 Seismic Playback Unit l SS-6336 Seismic Trigger l Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a seismic event data shall be retrieved from actuated instruments.

l The data shall be analyzed to determine the magnitude of the vibratory ground motion. A report shall be prepared and submitted to the NRC within ten days describing the magnitude, frt.quency spectrum, and resultant effect upon facility features important to safety.

4.1.4 MAINTENANCE ACTIVITIES Maintenance activities include corrective and preventive maintenance as well as periodic testing. Maintenance activities focus on the SFP cooling system, portions of the Component Cooling Water System (CCWS) and Service Water System (SWS) necessary to support SFP cooling, and components that support the emergency plan, fire protection plan, security plan i

or other licensed condition.

Revision 6 4.1-4 l

1

_______________.____________.___________________________________--__--___--_-_.--_-_U

4.1.5 ' ADMINISTRATIVE CONTROL OF SYSTEMS Plant systems and components that are no longer required to support defueled Plant operations will be deactivated. To prevent deactivated systems or components from affecting systems that are required to support defueled Plant operations, the following guidelines have been 1

incorporated into Plant procedures:

(1)_ Those systems required to support defueled Plant operations shall have physical isolation from systems or components that are deactivated. Isolation will be incorporated into Plant operating procedures.

.(2) Deactivated systems will be de-energized and isolation boundaries established to provide appropriate facility and personnel safety.

(3) Where practical, combustible material such as charcoal from filters, lube oils, f%.

Q electrohydraulic control fluids, etc., should be removed from systems or components not required for defueled Plant operations. I I

To ensure that the deactivation of systems or components does not result in a reduction of safety, deactivation of systems and components requires the performance of a screening to determine if a 10 CFR 50.59 evaluation is required.

Plant configuration control is implemented through administrative controls. These controls include:

1 L

(1) System drawings, as a minimum, should be revised to reflect changes in system

. lineups and equipment availability.

I 4.1-5 Revision 6 l

\

(2) Procedure revisions are performed as required when systems or components are deactivated.

I (3) Removal of components, installation of electrical jumpers, or cutting and capping l

piping systems will be performed in accordance with Plant design change procedures.

(4) Systems and/or components deactivated should be deenergized and labeled l

deactivated as appropriate.

Activities applicable to the safe storage of irradiated fuel as recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, are implemented and maintained by written procedures.

4 Revision 6 4.1-6

emergency makeup water source. Systematic guidance is provided by Plant procedures for leak identification and isolation.

The only requirement to assure adequate cooling for the spent fuel is to maintain the water level in the SFP so that the spent fuel elements are not exposed. All lines entering the SFP which could siphon the pool to Elevation 76 feet 7 inches or below (equivalent to approximately 10 feet above stored fuel assembly), are equipped with siphon breakers to limit SFP water loss to an elevation of 83 feet 11 inches. The SFP gates for access to the cask loading pit and refueling canal are equipped with inflatable seals to prevent level loss to these adjacent areas. For a postulated event wherein the SFP is drained to the Siphon Breaker level followed by failure of the SFP gates and subsequent spillage of SFP water to the fuel transfer canal and cask loading pit (initially assumed empty), SFP level would reach a minimum level of 76 feet 7 inches, or 10 feet above the top of the fuel assemblies. Additional discussion of the SFP design to prevent loss of level is contained in Section 3.2.2. Accident analyses for l p loss of SFP level are contained in Section 6.3.

V 4.31.3 Loss of Soent Fuel Pool Cooling As discussed in Section 4.3.1, a high temperature alarm is provided in the control room with a maximum setpoint of 135 F. Plant procedures direct actions to restore operation of SFP cooling. As stated in Section 4.3.1.2, the only requirement to assure adequate cooling for the spent fuel is to maintain the water level in the SFP such that the spent fuel elements are not exposed. Procedure guidance directs the restoration oflevel by the use of makeup to the pool in the event of level loss from boil off. Restoration of SFP level is discussed in Section 4.3.1.2, Loss of Spent Fuel Pool Level.

O 4.3-3 Revision 4 i

.1

1 l

4.3.1.4 High Soent Fuel Pool Level A level switch has been provided in the SFP for the purpose of transmitting high and low water level annunciation signals to the control room. Plant procedures are in place that i

)

provide guidance for the systematic determination and isolation of the water source. Overflow from the SFP is directed to the dirty waste drain tank.

4.3.1.5 Safety Criteria and Assurance The only requirement to assure adequate cooling for the spent fuel is to maintain the water level in the SFP so that the spent fuel elements are not exposed. The top of fuel seated in the spent fuel storage modules is located at Elevation 66 feet 7 inches. Dose calculations for fuel handling accidents assume a minimum of 23 feet of water above the top of spent fuel assemblies; therefore, SFP level less than Elevation 89 feet 7 inches requires operator action to restore level. This minimum required level also ensures dose rates at the SFP surface will be maintained <2.5 mrem /hr during fuel movements.

A SFP minimum boron concentration of 2000 ppm is assumed for certain design basis accidents to prevent inadvertent criticality. Sampling frequency and corrective actions to ensure boron concentration is maintained have been implemented.

A maximum temperature limit of 140*F was assumed for SFP bulk temperature. This limit ensures that for,the worst case loss of SFP cooling accident, SFP boiling will not occur within l 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br />, and that at least 10 days are available to establish makeup to the SFP before boil off would re' duce SFP level to 5 feet above the fuel assemblies. Maintaining 5 feet of water above the active portion of the fuel provides adequate radiation shielding to allow access for restoration of SFP level. Table 6.3-2 provides estimates of radiation dose rates for various SFP levels.

i Revision 6 4.3-4 h

4.3.2 FLFCTRICAL DISTRIBUTION During normal Plant operation, Plant electrical demands are provided from the 230-kV switchyard. The design and reliability of the offsite power source are discussed in

' Section 3.4.1.

Offsite power is normally supplied to the Plant distribution system via the startup transformers.

In the event of loss of offsite power, forced cooling capability would be lost. Without restoration C forced spent fuel cooling, boiling of the SFP would not occur for at least 88 hours0.00102 days <br />0.0244 hours <br />1.455026e-4 weeks <br />3.3484e-5 months <br />. SFP inventory is adequate to allow boiling for a minimum of 22.5 days without any source of makeup and still maintain > 10 feet of level above the top of the spent fuel assemblies. Makeup capability to the spent fuel coolant system independent of Plant power sources can be provided by the fi e main system (diesel-driven fire pump). A discussion of the fire main system is contained in PGE-1012, " Trojan Nuclear Plant Fire Protection Program."

4.3.3 SUPPCRT SYSTEMS l 1

Operation of the SFP cooling system is supported by the CCWS and SWS. Decay heat removed from the SFP cooling system is rejected to the CCWS via the SFP heat exchangers.

The SWS provides heat rejection of the CCWS to the Columbia River. Other Plant support i systems include ventilation, radiation monitoring, and radioactive waste systems, which are discussed in Chapter 5.

The CCWS interfaces with the SFP cooling system providing for heat rejection cf the spent fuel decay heat via a SFP cooling water heat exchanger. Normal operation consists of one .

componem cooling water pump in operation. As a result of Plant defueling, the in-service SFP heat exchanger is the only required component cooling water load. Component cooling 4.3-5 Revision 6 ,

L

water flow to the SFP heat exchanger may be throttled or secured to allow for maintaining SFP temperature between 40 F and 149 F. Loss of the CCWS can result in a loss of SFP cooling which is discussed in Section 4.3.1.3.

The loads and design basis of the SWS are discussed in Section 3.3.4. In addition to l providing heat rejection for the CCWS, the SWS provides a makeup water source for the SFP.

O I

Revision 4 4.34

5.3.2 DIRTY RADIOACTIVE WASTE SYSTEM O

-Tne DRWS, depicted in Figure 5.3-2, typically collects, processes, and monitors high particulate liquids.

1 5.3.2.1 Desicm no .

i

. The following design bases apply to the DRWS:

(1) To provide a means for collecting the liquid effluent from floor drains and sumps in the Containment Building, the Fuel Building, and the Auxiliary Building.

(2) To provide sufficient storage capacity for the maximum anticipated liquid flow from

~t he floor drains and sumps in the Containment Building, Fuel Building, and Auxiliary Building.

(3) To remove radioactive particulate by filtration and demineralization.

(4) To reduce the concentration of radioactivity in discharge streams to a level ti.at, when -

combined with other liquid effluents, will permit release to unrestricted areas within the guidelines of Appendix I to 10 CFR 50, and limit the dose to r.ny organ to 5;.5 mR during any calendar year (U.

(5) To provide storage capability in the discharge path for monitoring the liquid prior to release and for recycling those batches that do not meet the design objectives.

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5.3.2.2 System Description The primary sources of liquids collected in this system are floor drains and egoipment drains from the Containment Building, Fuel Building, and Auxiliary Building.

The DRWS consists of the following components:

(1) Reactor cavity sump.

(2) Reactor cavity sump pump.

(3) Two containment sumps.

l (4) Two Auxiliary Building sumps (ABS).

l (5) Two Auxiliary Building sump pumps.

l (6) Auxiliary Building passageway sump (ABPS).

l (7) Auxiliary Building passageway sump pump.

l (8) Dirty waste drain t.ank (DWDT).

l (9) Two DWDT pumps.

l (10) Two dirty waste bag 51ters.

l (11) Interconnecting piping, valves, and instrumentation.

O Revision 6 5.3-6

The reactor cavity sump collects leakage, washdown water, and condensation in the lowest A

Q section of the Containment Building (Elevation 16 feet). The reactor cavity sump pump (air operated) is used to transfer the liquid to the "A" (South) containment sump.

I

- The Auxiliary Building sumps "A" and "B" collect drainage from floor and equipment drains in the rooms at Elevation 5 feet. The sumps are covered and are vented to the vent collection header. One Auxiliary Building sump pump in each sump is used to transfer the sump contents to the DWDT.

i The Auxiliary Building passageway sump collects drainage at Elevation 5 feet that does not go.

to the ABSs. The ABPS pump is used to transfer the sump contents to the DWDT (normally),

or to the CWRT or TWMT in the clean radioactive waste system.

The Auxiliary Building sump pump, and ABPS pump will normally automatically operate l when accumulated drainage reaches preset sump levels.

The DWDT receives input from the DRWS sumps and directly from various tank, equipment and floor drains through air-cperated inlet valves. The tank is evenly divided into two sections. The tank sections are connected by an unvalved overflow b that has a 3-foot loop seal drain cennection to the ABPS. This provides tank overpressure protection while preventing the possible spread of contamination. In addition, the tank is vented to the vent collection header to prevent gas buildup.  :

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Each DWDT section has a sparger with an air cor.nection that is used to thoroughly mix tank contents for representative sampling and to prevent settling of particulate matter.

Two DWDT pumps take a suction on the tank sections and allow the DWDT contents to be processed and transferred. The DWDT contents are normally processed through the DWDT bag filters and the demineralizers (see Section 5.3.1) to the treated waste monitor tanks.

Connections also exist to process dirty radwaste through the steam generator blowdown demineralizers.

l 5.3-7 Revision 6 l-

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The dirty waste bag filters are normally used to remove particulate matter from the DWDT pump discharge flow. Differential pressure indication and alarm are provided to alert the ,

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operator to excessive filter clogging which will require filter replacement.

5.3.2.3 Design Evaluation The DRWS does not perform any safety functions for the permanently defueled Plant-condition. The system collects waste water from various sources and processes it for discharge. The DRWS has adequate capacity to process waste from the small number of input sources in opration with the Plant in the permanently defueled condition.

Administrative controls are established to require further processing when necessary to ensure effluent releases are within allowable limits.

4 Revision 6 5.3-8

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' 7.1.2.2 Supoorting Organi7mtiom j

The Supporting Organizations consist of four main groups supporting Plant activities:

Engineering / Decommissioning, Plant Support and Technical Functions, and Nuclear Oversight. The following summarizes the responsibilities and authority of major supporting organization positions:

General Manager. Engineering /DecommiccSning . reports to the General Manager, Trojan i Plant, and is responsible for engineering assistance, decommissioning, fire protection, and-l ' design control of Trojan. Under the direction of the General Malmger, Engineering / Decommissioning, engineering support is provided for safe, efficient operation of plant equipment, drawing and design document preparation, and plant modifications.

t Manager. Decommiuioning PInnning - responsible for the development of a l decommissioning plan for the Trojan Nuclear Plant including the transition to dry fuel

) storage.

I Manager. Engineering - responsible for engineering assistance, modifications, l

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maintaining Plant configuration data and documentation, oversight and direction of the Fire Protection Program and Civil / Seismic Program. Also responsible for providing L technical support for Plant Operations and Maintenance.

Proiect Manager Comnorent Removal Proiects - responsible for component removal activities as they relate to facility decommissioning.

General Manager. Plant Supoort and Technien1 Functions - Reports to the General Manager, Trojan Plant, and has responsibility for the Plant Support and Technical Functions organizations, purchasing and cost control, security, training, licensing, compliance and l commitment management, and the overall administration and maintenance of records.

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Manager. Nuclear Security - reports to the General Manager, Plant Support, and is O

respons? 'e for the administration of the Plant Security Organization, implementation of the secury program, and the Fitness-for-Duty Program. This responsibility includes interaction with state and federal regulatory agencies; communication and coordination with local law enforcement agencies; direct supervision of the security staff; administration of the contract with the security contractors; selection, training and staffing of the security organization; administering the security screening programs for personnel authorized unescorted access to the Plant or to Safeguards Information; preparation of procedures required to implement the security program; approval of security-related Plant Modification Requests (PMRs).

Manager. Licensing. Comoliance. and Commitment Management - responsible for all activities required to maintain the permits and licenses required for the Plant including producing, maintaining, and interpreting licensing documents.

l Manager. Cost Control - responsible for cost control and procurement of equipment l and services.

General Manager. Nuclear Oversight - reports to the General Manager, Trojan Plant, and is responsible for the QA Program, for audits that are carried out to verify compliance with the QA Program, for evaluating the effectiveness of the QA Program for each area that is audited and for supporting line management in event analysi:;, review, and performance monitoring of activities that affect the safe and reliable operation of the facility. He has the authority and independence to identify quality problems; initiate, recommend, or provide solutions to quality problems through designated channels; and verify implementation of solutions to quality problems. He has the authority and responsibility to initiate stop work orders to responsible management, as necessary, for any condition adverse to quality.

Revision 6 7.1-4

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