ML20209C653

Rev 8 to Defueled SAR, for Trojan Nuclear Plant

ML20209C653
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/06/1999
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20209C651	List: ML20209C648 ML20209C653
References
NUDOCS 9907120162
Download: ML20209C653 (145)

Text

!

. l TABLE OF CONTENTS

(

\ -

DEFUFI Rn SAFETY ~ ANALYSIS REPORT CHAPTER 1.0 INTRODUCTION AND

SUMMARY

Section Title lagC_

1.0 INTRODUCTION

AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . 1.0-1 1.1 Intradnetion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 1.2 General Plant Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 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 Idantification of Agents and Contractors . . . . . . . . . . . . . . . . . . . I 1.3-1

-1.3.1 Design of the Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.4 Exclusions from and Exemntions to certain Parts of Title 10 of the Code of Federal Reaulations (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 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 Powu 14cilities Operating Prior to Januarv 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 Pievention Measures for Light-Water Nuciear Power Reac^;ars ....... 1,4-3 1.4.1.9 10 CFR 50.61, Fracture Toughness Requiremw.ts for  :

Protection Against Pressurized Thermal Shock Events . .. . . . . . 1.4-3 i Revision 7 9907120162 990706 PDR ADOCK 05000344 W. PDR i

CHAPTER 1.0 INTRODUCTION AND

SUMMARY

h Section Title Page 1.4.1.10 10 CFR 50, Appendix G, Fracture Toughness Requirements . . 1.4-3 1A.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, Mzking of Reports . . 1.4-5 1.4.1.15 10 CFR 70.24, Criticality Accident Requiretiants . . . . . . . . . . 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 ............................... 1.4-6 1.4.2.2 10 CFR 50.54(y), Conditions 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 Revision 8 ii O

CHAPTER 2.0 O SITE CHARACTERISTICS .

Section Title _Eage_

2.0 SITE CHARACTERISTICS . . . . . . . . . . . . . . . . . . . . . . ... 2.0-1 2.1 Geography and Demography . . . . . . . . . . . . . . . . . . . . . . . . . . 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 Boundaries for Establishing Effluent Release Limits . . . . . . . . . 2.1-3 2.1.2 Exclusion Area Authority and Control . . . . . . . ............ 2.1-3 2.1.2.1 Au tho rity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-3 2.1.2.2 Exclusion Area Activities Unrelated to Plant Operation . . . . . . . 2.1-5 2.1.2.3 Arrangements for Traffic Control . . . . . . . . . '. . . . . . . . . . . . 2.1-6 2.1.3 Population Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-6 2.1.3.1 Population Within 10 Miles . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-7 j 2.1.3.2 Population Between 10 and 50 Miles . . . . . . . . . . . . . . . . . . . 2.1-8 2.1.3.3 Transient Population . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-9 2.1.3.4 Low-Population Zone ............................ 2.1-10 2.1.3.5 Population Center . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-10 2.1.4 Uses of Adjacent Lands and Waters . . . . . . . . . . . . . . . . . . . . . 2.1-10 2.2 Nearby Industrial. Transoortation 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 Pipeline s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-5 2.2.2.3 Waterways ................................... 2.2-5 2.2.2.4 A irpo rts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 2.2.3.4 Ship Collision with Intake Structure ................... 2.2-20 2.2.3.5 Oil or Corrosive Liquid Spills in River . . . . . . . . . . . . . . . . 2.2-21 iii Revision 8

CHAPTER 2.0 SITE CHARACTERISTICS Section Title Page 2.2.3.6 Cooling Tower Collapse . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2-22 2.3 M eteo rol ogy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 . . . . . . . . . . . . . . 2.3-5 2.3.4 Diffusion Estimates . . . . . . . ........................ 2.3-6 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 l 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 LocaiIntense Prxipitation .................. 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 7 iv O

l

[

[

CHAPTER 3.0 (q FACILITY DESIGN

%.) .

Section Title Page 3.0 FACILITY DESIGN . . . . . . . . . .................... 3.0-1 3.1 Summarv ...................................... 3.1-1 I 3.1.1 Conformance with NRC General Design Criteria . . . . . . . . . . . . . 3.1-1 3.1.2 Classification of Structures, Components and Systems ......... 3.1-7 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 Level (Flood) Design . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-13 3.1.5 Missile Protectic . Criteria ........................... 3.1-14 3.1.5.1 Missile Selection and Description . . . . . . . . . . . . . . . . . . . . . 3.1-14 3.1.6 Seismic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 Soent 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 Design 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 vii Revision 7

CHAPTER 3.0 I iCILITY DESIGN Section Title Page 3.2.2.2 System 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 Au x ili a ry Sy s tem s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.1 Fuel Handling System . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 3.3-1 3.3.1.1 Design Bases . . . . . . . .......................... 3.3-1 3.3.l.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-2 3.3.1.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 3.3-4 3.3.2 Modular SFP Cooling and Cleanup System . . . . . . . . . . . . . . . . . 3.3-5 3.3.2.1 Des ign Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-5 3.3.2.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-6 3.3.2.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-8 h

3.3.3 Deleted 3.3.4 Service Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-9 3.3.4.1 Design Bases . . . . . . ....................... ... 3.3-9 3.3.4.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-10 3.3.5 Compressed Air System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-11 i 3.3.6 Boric Acid Batch Tank . . . . ......................... 3.3-11 l 3.3.7 Deleted 3.3.8 Equipment and Floor Drain Systems . . . . . . . ....... .... 3.3-12 3.3.9 Plant Discharge and Dilution Structure . . . . . . . . . . . . . . . . . . . 3.3-13 i

3.3.10 Deleted Revision 8 viii 1

)

i 9

)

i l

j CHAPTER 3.0 f FACILITY DESIGN Section Title Page 3.3.11 Fire Protection System and Program . . . . . . . . . . . . . . . . . . . . . 3.3-1.,

3.3.12 Control Room Habitability . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-13 3.3.13 Seismic Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-14 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 Analy s is . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-4 3.5 Comnliance with NRC Reentatory Guides ................. 3.5-1 C

3.6 References . . . . . . . . . ............................ 3.6-1

)

l e

CHAPTER 4.0 OPERATIONS Saction Title Page 4.0 OPERATIONS . . . . . . . . . . . . ...................... 4.0-1 4.1 Ooeration 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 Functicnal 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 Suonort 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 l

4.4 Control Room Area . . . ............................ 4.4-1 4.5 References .................................... 4.5-1 l Revision 7 x 0

CHAPTER 5.0 m '

~ RADIATION PROTECTION U

Section Title Page 5.0 RADIATION PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . 5.0-1

. 5.1 Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.1-1 5.2 Offgas Treatment and Ventilation . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.1 Deleted l

5.2.2 Containment Ventilation System . . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.2.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.2.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-2 5.2.2.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-3 5.2.3 Fuel and Auxiliary Building Ventilation System . . . . . . . . . . . . . . 5.2-3 5.2.3.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-4 5.2.3.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-4 5.2.3.3 Design Evaluation . , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-5 5.2.4 Radwaste Processing Building Ventilation System . . . . . . . . . . . . 5.2-5 5.2.4.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-6 5.2.4.2 S) stem Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-6 5.2.4.3 Design Evaluation . . . . . . . . ,. . . . . . . . . . . . . . . . . . . . . . . 5.2-7 5.2.5 Modular Spent Fuel Pool Cooling System Cooling Air . . . . . . . . . 5.2-7 5.3 Liauid Waste Treatment and Retention ....... ........... 5.3-1 5.3.1 Des ig n B ases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-1 5.3.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-2 5.3.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-4 5.4 Solid Wa ste s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-1 0 xi Revision 8 1

CHAPTER 5.0 RADIATION PROTECTION -

Section Title Page 5.4.1 Des ign Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-1 5.4.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-2 5.4.2.1 Spent Resin Transfer System . . . . . . . . . . . . . . . . . . . . . . . . 5.4-2 5.4.2.2 Filter Handling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-2 5.4.2.3 Solid Wastes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-2 5.4.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-3 5.5 Process and Effluent Monitoring Systems . . . . . . . . . . . . . . . . . . 5.5-1 l 5.5.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-1 5.5.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-2 5.5.2.1 Liquid Monitoring Systems . . ................... .. 5.5-3 5.5.2.2 Gas Monitoring Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-4 5.5.2.3 Analytical Procedures ............................ 5.5-7 5.5.2.4 Calibration and Maintenance . . . . . . . . . . . . . . . . . . . . . . . . 5.$-8 1

5.5.3 Effluent Monitoring and Sampling . . . . . . . . . . . . . . . . . . . . . . 5.5-9 5.5.4 Process Monitoring and Sampling ...... ............... 5.5-9 l

l 5.6 Radiation Protection Program ......................... 5.6-1 5.6.1 Radiation Protection Design Features . . . . . . . . . . . . . . . . . . . . 5.6-1 i 5.6.1.1 Shielding, Radiation Zoning and Access Control . . . . . . . . . . . 5.6-1 5.6.1.2 Plant Ventilation Systems . . . . . . . . . . . . . . . . . . . . ..... 5.6-1 5.6.1.3 Area Radiation Monitoring Instrumentation .......... ... 5.6-1 5.6.2 Equipment, Instrumentation and Facilities . . . . ......... . 5.6-2 5.6.2.1 Radiation Protection Facilities . . . . . ............... . 5.6-3 5.6.2.2 Radiation Protection Instrumentation .. ...... .. .... 5.6-4 5.6.3 Procedures ..... .... ........ ........... ... . 5.6-6 5.6.3.1 Control of Personnel Radiation Exposure . . . . . . . . . . ... 5.6-6 5.6.3.2 Personnel Dosimetry . . . . . . ............. ....... 5.6-8 Revision 8 . xii 9

l

LIST OF FIGURES -

DEFUFI FD SAFETY ANALYSIS REPORT Numher Title 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 117 feet 0 inch 3.1-19 3-D Model for the Fuel Building Steel Superstructure 3.2-1 Fuel Building - Plan Elevation 45' 3.2-2 . Snent Fuel Pool - Typical Details 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' -

3.2-7. 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 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, Reacior and Auxiliary Buildings - Plan Below Ground Floor 3.2-16 Equipment Location, Reactor and Auxiliary Buildings - Plan Operating Floor, Elevation 45' 3.2-17 Equipment Location, Reactor and Auxiliary Buildings - Plan Elevation 61' 3.2-18 Equipment Location, Reactor and Auxiliary Buildings - Plan Elevation 77' 3.2-19 Equipment Location, Reactor and Auxiliary Buildings - Plan Operating Floor and Above 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 xxiii Revision 7

LIST OF FIGURES DEFUEI ED SAFETY ANALYSIS REPORT O:

Number Title 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 3.3-4 Modular Spent Fuel Pool Cooling and Cleanup System ,

4.2-1 Spent Fuel Storage Pool

~

l 4.2-2 Deleted l 5.2-1 Deleted 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 Demineralizer Building Ventilation Exhaust System

~

5.3-1 Liquid Radioactive Waste System ,

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 8 xxiv

LIST OF EFFECTIVE PAGES DEFUFI FD SAFETY ANALYSIS REPORT Section Effective Pages Revision Title Page N/A Rev.O Table of Contents i Rev.7 Table of Contents ii and iii Rev.8 Table of Contents iv through vii Rev.7 Table of Contents viii and ix Rev.8 Table of Contents x Rev.7 Table of Contents xi and xii Rev.8 4

Table of Contents xiii through xxiii Rev.7 l Table of Contents xxiv through xxix -

Rev.8 1.0 1.0-1 Rev.0 1.1 1.1-1 Rev.0 1.2 1.2-1 and 1.2-2 Rev.8 1.3 1.3-1 Rev.0 1.4 .1.4-1 through 1.4-4 Rev.4

-Q(/

1.4 1.5 1.4-5 and 1.4-6 1.5-1 Rev.8 Rev.8 l Figure 1.1-1 N/A Rev 0 l 2.0 2.0-1 Rev.0 2.1 2.1-1 through 2.1-12 Rev.8

.2.2 2.2-1 through 2.2-18 Rev.8 2.2 2.2-19 Rev.0

. 2.2 2.2-20 Rev.6 2.2 2.2-21 Rev.7 2.2 2.2-22 Rev.0 2.3 2.3-1 and 2.3-2 Rev.0 2.3 2.3-3 Rev.7 2.3 2.3-4 Rev.0 2.3 2.3-5 through 2.3-10 Rev.7 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 and 2.4-33 ,

Rev.7

\ ..

xxv Revision 8

LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision 2.4 2,4-34 and 2.4-35 Rev.0 2.4 2.4-36 Rev.6 2.5 2.5-1 Rev.7 2.5 2.5-2 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 Rev.0 2.6 2.6-1 through 2.6-10 Rev.8 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.8 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 3.2-37 Rev.5 3.2 3.2-38 Rev.7 3.2 3.2-39 through 3.2-41 Rev.5 f 3.3 3.3-1 Rev.8 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 3-9 Rev.7 3.3 3.3-10 through 3.3-14 Rev.8 3.4 3.4-1 Rev.O 3.4 3.4-2 and 3.4-3 Rev.4 i

3.4 3.4-4 Rev.6 3.5 3.5-1 Rev.0 l 3.6 3.6-1 through 3.6-3 Rev.8

! Tables 3.1-1 and 3.1-2 N/A Rev.0

( -

I Revision 8 xxvi 1

I

1 LIST OF EFFECTIVE PAGES f3 U -

DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision Table 3.1-3 N/A Rev.6 Tables 3.1-4 through 3.1-7 N/A Rev.O Tables 3.2-1 through 3.2-3 N/A Rev.O Table 3.2-4 N/A Rev.1 Table 3.2-5 N/A Rev.7 i Table 3.51 Sheets 1 through 35 Rev.8 Figures 3.1-1 through 3.1-19 N/A Rev.O Figures 3.2-1 N/A Rev.7 Figures 3.2-2 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.0 Figure 3.3-4 N/A Rev.7 4.0 4.0-1 and 4.0-2 Rev.7 4.1 4.1-1 and 4.1-2 Rev.7 4.1 4.1-3 and 4.1-4 Rev.8 s - 4.1 4.1-5 and 4.1-6 Rev.6 4.2 4.2-1 through 4.2-3 Rev.8 j 4.3 4.3-1 through 4.3-3 Rev.7 l 4.3 4.3-4. Rev.6 l 4.3 4.3-5 Rev.7 l 4.4 4.4-1 Rev.7 4.5 4.5-1 Rev.8

. Table 4.3-1 N/A Rev.5 Figure 4.2-1 N/A Rev.4 5.0 5.0-1 Rev.0 5.1 5.1-1 Rev.0 5.2 5.2-1 through 5.2-7 Rev.8 5.3 5.3-1 through 5.3-4 Rev.8 5.4 5.4-1 through 5.4-3 Rev.8 5.5 5.5-1 Rev.0 5.5 5.5-2 and 5.5-3 Rev.7 5.5 5.5-4 through 5.5-9 Rev.8 5.6 5.6-1 Rev.1 5.6 5.6-2 Rev.3 xxvii Revision 8

n LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision 5.6 5.6-3 Rev.0 5.6 5.6-4 through 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 and 5.9-2 Rev.8 Table 5.5-1 N/A Rev.7 Figures 5.2-2 through 5.2-4 N/A Rev.5 Figure 5.2-5 N/A Rev.O Figure 5.2-6 N/A Rev.7 Figure 5.2-7 N/A Rev.5 Figure 5.3-1 N/A Rev.8 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.8 6.3 6.3-1 through 6.3-6 Rev.7 6.4 6.4-1 through 6.4-3 Rev.8 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 Rev.3 7.1 7.1-1 Rev.4 7.1 7.1-2 Rev.3 7.1 7.1-3 Rev.6 7.1 7.1-4 Rev.8 7.1 7.1-5 Rev.3 7.2 7.2-1 and 7.2-2 Rev.3 7.2 7.2-3 through 7.2-6 Rev.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 7.5-1 Rev.4 Revision 8 xxviii ,

LIST OF EFFECTIVE PAGES O DEFURTFD SAFETY ANALYSIS REPORT Section Effective Pages Revision 7.6 7.6-1 Rev.3 8.0 8.0-1 Rev.0 9.0 9.0-1 Rev.0 I

1 1

1 O ,

e l

O xxix Revision 8 l

l

O O

O

l l

J J,.2 GENERAL PLANT DESCRIPTION

' ("\ ,

V.

1 The Trojan Nuclear Plant site consists of approximately 623 acres located in Columbia County i in NW Oregon on the Columbia River at River Mile 72.5 from the mouth. The distance from the reactor site to the nearest site boundary on land is 2172 ft. Major structures on the site include the Containment, Turbine Building, Auxiliary Building, Fuel ' Building, Control Building, and a single natural draft cooling tower.

The Trojan Independent Spent Fuel Storage Installation (ISFSI) area is within the Exclusion l

Area of the Trojan Nuclear Plant. The ISFSI access controlled area of the site is subject to l 10 CFR 72 License No. SNM-2509, and is not part of the site licensed under 10 CFR 50 l (License No. NPF-1). Characteristics of the ISFSI site are described in the Trojan ISFSI l Safety. Analysis Report, FGE-1069.

l The town of St. Helens, Oregon, the county seat of Columbia County, is located approximately 12 miles SSE of the site. The town of Rainier, Oregon, is located approximately 4 miles NNW and the town of Kalama, Washington, is approximately 3 miles SE of the site. There are three small unincorporated communities within a 5-mile radius of the site: Prescott, Oregon,' located approximately 1/2 mile N of the site; Goble, Oregon, located approximately 1-1/2 miles SSE of the site; and Carrolls, Washington, located approximately 2-1/2 miles NNE of the site.

1 1.2.1 DESIGN CRITERIA The principal design criteria for the Trojan Nuclear Plant are those fundamental architectural and engineering design objectives established for the Facility. The basis for development and selection of the design criteria used in this Facility were those which: (a) provide protection of public health and safety, (b) provide for reliable and economic Facility performance, and (c) provide an attractive appearance, i 1.2-1 Revision 8 f

The essential systems and components of the Facility are designed to enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might .

be imposed by natural phenomena. The designs are based on the most severe of the natural phenomena recorded for the vicinity of the site, with margin to account for uncertainties in historical data.

1.2.2 FUEL HANDLING SYSTEM The Facility was designed to handle spent fuel under water from the time it leaves the reactor vessel until it is placed in casks for shipment from the site, although PGE is prohibited from moving fuel into the containment. Underwater transfer of spent fuel provides an optically transparent radiation shield, as well as a reliable source of coolant for removal of decay heat.

1.2 3 RADIOACTIVE WASTE TREATMENT SYSTEMS The radioactive waste treatment systems provide equipment necessary to collect, process, monitor, and discharge radioactive liquid, gaseous and solid wastes that are produced at the Facility.

Liquid wastes potentially containing radioactive material are collected and monitored. Prior to discharge, equipment is provided for filtering the liquid as required. Also, connections exist for a demineralizer if filtration alone is not sufficient to meet discharge requirements. The treated water from the filters or demineralizers may be recycled for use in the Facility or may be discharged to the Columbia River. The miscellaneous dry waste, spent demineralizer resins and spent filters are shipped from the site for ultimate disposal in an authorized location.

l Gaseous wastes are discharged to the environment after filtration to keep the offsite dose within prescribed limits.

Revision 8 1.2-2

I i

1 1.4.1.14 ' IJ CFR 50.71(et Maintenance of Record. Makina of Reoorts C/

This regubtion 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  !

down, and altered the rule to recognize the limited safety importance of Safety Analysis Updates for shutdown plants. The modified final rule is, therefore, now applicable to Trojan.

I 1.4.1.15 10 CFR 70.24. Criticalitv Accident Reauirements '

By letter dated March 24,1993 from S. H. Weiss (NRC) to J. E. Cross (PGE), the staff has 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 l assurance against the occurrence of an accidental criticality exists that this regulation no longer applies to Trojan in its permanently shutdown and defueled condition. However, some l conditions on which'the NRC based this notification are invalidated during spent fuel loading l 1

into PWR Baskets. Rather than request an exemption from 10 CFR 70.24, PGE has instead l chosen to comply with 10 CFR 50.68(b), as provided for by 10 CFR 70.24(d). l 1.4.2 EXEMPTIONS TO 10 CFR RELATED TO THE PERMANENTLY DEFUFT FD CONDITION l I

There are also certain 10 CFR regulations to which the NRC has granted specific exemptions to PGE for the operation of the Trojan Facility.

l O

v 1.4-5 Revision 8

1.4.2.1 10 CFR 50.54(o) and Appendix J, Primary Reactor Containment Lenkage 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.

1.4.2.2 10 CFR 50.54(v). 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.

1.4.2.3 10 CFR 50.54(q) and Certain Sections of 10 CFR 50.47. " Emergency Plans." and 10 CFR 50. Apoendix E. " Content of Emergency Plans" By letter dated September 30,1993, from M. T. Masnik (NRC) to J. E. Cross (PGE), the O

NRC authorized the Trojan facility to discontinue offsite emergency preparedness activities and reduce the scope of onsite planning.

4 1

.1 Revision 8 1.4-6 l

1.5 MATERIAL INCORPORATED BY REFERENCE Certain program manuals and topical reports have been incorporated into the DSAR by reference and are listed in the last section of each chapter. The reports include topical reports written by PGE as well as by Westinghouse, Bechtel and other orgamzations.

Some documents that are incorporated by reference continue to be updated to assure a.st the information presented is the latest available. These documents include those listed below:

(1) PGE-1060, " Permanently Defueled Emergency Plan."

~

(2) PGE-1012. " Fire Protection Plan."

(3) PGE-1017, " Security Plan."

(4) PGE-1020, " Report on Design Modifications for the Trojan Control Building."

(5) PGE-8010, " Nuclear Quality Assurance Program."

(6) PGE-1052, " Quality-Related List Classification Criteria for the Trojan Nuclear Plant."

(7) PGE-1021, "Offsite Dose Calculation Manual."

(8) PGE-1057, " Trojan Nuclear Plant Certified Fuel Handler Training Program."

(9) PGE-1024, " Trojan Nuclear Plant Security Force Training and Qualification Plan (Defueled Condition)."

(10) PGE-1%1, " Decommissioning Plan."

(11) License Change Application (LCA 237), " Spent Fuel Cask Loading in the Fuel l Building," transmitted by PGE letter, VPN-005-99, dated January 7,1999. l (12) LCA 246, " Spent Fuel Cask Loading in the Fuel Building - Contingency Fuel l Unloading to the Fuel Pool," transmitted by PGE letter, VPN-006-99, dated l January 27,1999. l 1.5-1 Revision 8

- )

(

2.1 GEOGRAPHY AND DEMOGRAPHY -

O .

The Trojan site was originally selected to minimize hazards to the general public. The site environs have low population densities and minimal usage for such activities as farming and recreation. Some of the site characteristics associated with the Trojan site selection for the operational phase of the facility remain applicable to defueled condition and the storage of irradiated fuel. This chapter provides discussion of those site characteristics applicable to defueled operation.-

2.1.1 SITE LOCATION AND DESCRIPTION 2.1.1.1 Specification of Location The Trojan Plant site is in Columbia County, Oregon, and lies along the bank of the Columbia River at approximately River Mile 72.5,42 miles north of Portland. The specific geographic A location of the site is 46* 02' 25" N latitude and 122 53' 03" W longitude. In the Universal O

Transverse Mercator coordinate system, the site location is 5098352 meters N by 509000 meters E, and in the Oregon North Zone Lambert Coordinate, the site location is 874375 N by 1394615 E.

t .

, The nearest incorporated communities are Rainier, Oregon, approximately 4-1/2 miles northwest; and across the Columbia River, Kalama, 3 miles southeast, and Longview, approximately 6 miles northeast. Within a 5-mile radius of the site are three small unincorporated communities with a total population of less than 2000: Prescott, Oregon,1/2-mile north; Goble, Oregon,1-1/2 miles southeast; and Carrolls, W :n,2-1/2 miles northeast.

' Other than the Columbia River and tributaries, there are no nearby natural geographic features of 1 prominence offsite. The Kalama River joins the Columbia at River Mile 73.1, about 1/2-mile upstream on the bank opposite the site. Similarly, the confluence of the Cowlitz and Columbia O 2.1-1 Revision 8

Rivers is about 4-1/2 miles downstream at River Mile 68. Onsite, however, are the 499 feet natural draft cooling tower, which rises 589 feet above mean sea level (MSL), an approximately 26-acre man-made reflecting lake and an approximately 28-acre recreational lake.

2.1.1.2 Site Area Map The Trojan Nuclear Plant site is an approximately 635-acre tract of land owned in fee by Portland General Electric Company (PGE) in Sec. 35 and 36, T. 7 N., R. 2 WWM, and in Sec. I and 2, T. 6 N., R. 2 WWM, Columbia County, Oregon. The tract is all-inclusive of individual and separate parcels as described in the following deed records on file in Columbia County: BK 168, Pages 13 and 14,22,23 to 26 inclusive, 81 to 83 inclusive,117 to 121 inclusive; BK 171, Pages~ 935 and 936; and BK 174, Page 436.

The exclusion area, is defined in 10 CFR 100.3(a). The exclusion area boundary coincides with the site boundary on the Oregon side of the river and extends across the Columbia River to the east where the Washington shore of the river forms the eastern boundary.

l The Trojan Independent Spent Fuel Storage Installation (ISFSI) area is within the Exclusion Area l of the Trojan Nuclear Plant. The ISFSI access controlled area of the site is subject to 10 CFR 72 l License No. SNM-2509, and is not part of the site licensed under 10 CFR 50 (License No.

l NPF-1). Characteristics of the ISFSI site are described in the Trojan ISFSI Safety Analysis l Report, PGE-1069.

The major physical facilities are grouped approximately in the geographic center of the exclusion ,

area. The cratcr of the containment lies 2172 feet due south of the nearest point on the exclusion area perimeter, approximately 3175 feet from the nearest point to the west, and approximately 4400 feet from the closest southerly point. Because of its irregular shape, the exclusion area comes within about 4200 feet of the containment at two points approximately southwest of the building. The containment center is located about 2200 feet from the nearest point of approach of Revision 8 2.1-2 h

the eastern boundary of the exclusion area, and 400 feet from the Oregon bank (mean low water) of the Columbia River.

2.1.1.3 - Boundaries for Establishine Effluent Relenke I imits As pointed out in the preceding section, the site boundary and exclusion area coincide on land on the Oregon side of the Columbia River.

Programmatic requirements for specifications applying to releases of radioactive material in gaseous effluents are given in the Offsite Dose Calculation Ma'nual. Doses and release limits have been evaluated at the site boundary and at off-site locations ~of actual exposure.

2.1.2 EXCLUSION AREA AUTHORITY AND CONTROL '

2.1.2.1 Authority O

The site boundary (owned in fee) extends to mean low water in the southern part of the site and to mean high water in the northern part of the site. By written agreement with the State of Oregon, who is owner of the submerged and remainder of submersible lands in the river at the site, PGE has control of the uses of such areas out to a line at approximately -20 feet MSL.

Beyond this line the U. S. Coast Guard has jurisdiction over river operations.

The provisions of the tidelands agreement with the State of Oregon include the following conditions:

(1) RESTRICTED USE: Residential use of and overnight camping in the exclusion area

- shall be prohibited by the State. Nonresidential activities and uses unrelated to the operation of the Company's adjacent reactor shall be permitted only when no significant t

O 2.1-3 Revision 8

hazard to public health and safety exists and then only under appropriate limitations as provided in this Agreement.

(2) USE LIMITATIONS: In managing its lands within the above-described exclusion area and in sales and leases made with respect to such lands, the State will insert in each Deed, Lease, Easement, Permit or other instrument granted a provision to the effect that the lands affected are within an exclusion area as that term is defined by the Nuclear Regulatory Commission; that such lands are subject to safety regulations established by the State, the Nuclear Reylatory Commission, and the Company with which the grantee shall comply; and that the Company has a right to remove or order the removal of all persons and their property therefrom in compliance with said safety regulations.

(3) ' EMERGENCY PROCEDURES: In the event of an emergency or threat thereof, which may affect public health or safety, the State grants the Company the right to enter upon its lands within the exclusion area and to remove persons and property therefrom. The State also grants the Company the right to proclaim safety regulations affecting persons and property occupying State owned !arrJs within the exclusion area, which regulations upon approval by the State shall be made by the State a part of every Deed, Lease, Easement, Permit or other instrument issued by it as previously provided. Such rules and regulations when so adopted shall also become a part of the State's management

, policy for the administration of such lands.

l Similar conditions were negotiated with the Burlington Northern Railroad (1978) and, l subsequently, with Portland & Western Railroad (1999). Mineral rights not part of the original land purchase at Trojan have been subsequently bought in fee by PGE.

Revision 8 2.1-4 o J

2.1.2.2 Exclusion Area Activities Unrelated to Plant Ooeration The Trojan Independent Spent Fuel Storage Installation (ISFSI) area is within the Exciusion Area l of the Trojan Nuclear Plant. Characteristics of the ISFSI site and a description of ISFSI activities l are provided in the Trojan ISFSI Safety Analysis Report, PGE-1%9. l The basic reference for activities and facilities wkhin the exclusion area is "The Trojan Nuclear Power Plant: Master Plan Prepared by Lawrence Halprin & Associates". In this document recreational uses (daytime only) are identified as hiking, picnicking, swimming, fishing and nature observation.

Doses 'that could be anticipated in the event of a (DBA) have been estimated and shown to be below the limits of 10 CFR 100.

Emergency planning is discussed further in the Trojan Permanently Defueled . Emergency Plan (PGE-1060).

With the permanent shutdown and defueling of the facility activities allowed in the exclusion area will be expanded. Accident analysis have demonstrated that operations and design basis accidents (DBA) do not result in radiological consecptences which would exceed 10 CFR 100 guidelines.

. As a result, buildings within the exclusion area but outside the protected area, may be made available for commercial use. Prior to commencement of commercial activities within the exclusion area an evaluation of the proposed activity will be performed to ensure that the accident analyses for the facility remain bounding. Leases issued to commercial users of these buildings will limit activities to ensure postulated event and accident analyses remain bounding.

1 2.1-5 Revision 8

2.1.2.3 Arrangements for Traffic Control The exclusion area is traversed in a generally north-south direction by a highway, railroad and the Columbia River. The highway, U.S. Highway 30, is a two-lane roadway carrying moderate passenger and freight traffic between communities along the Columbia River. The highway l right-of-way passes within 2800 feet of the containment. The rail right-of-way is operated by l Portland & Western Railroad, which operates an average of two freight trains per day along this line. The track passes approximately 600 feet from the containment. All railroad property within the exclusion area is " operating" property, i.e., not available for lease or other use.

The eastern portion of the exclusion area is a stretch of the Columbia River, a major waterway both for commerce and recreation, which also serves as the border between Oregon and m

Washington. Approximately six oceangoing commercial vessels pass the site on a typical day ,

In the event of an emergency that could result in a hazard to the general public, several raeasures can be initiated to control traffic through and access to the exclusion area. Members of the general public mr. king recreational or other casual use of the nonrestricted portions of the exclusion area can be evacuated. Roadblocks can be established as required at the junction of the highway with the northern and southern boundaries of the exclusion area, thereby restricting road travel through the site. Railroad traffic can be restricted over the rail right-of-way through the exclusion area. River traffic can be closed off from the portion of the river within the exclusion area.

( 2.1.3 POPULATION DISTRIBUTION Population and population distributions within a 50-mile radius of the site are evaluated in this section using the 1970 census data" and projected to the year 1980 by decades. These figures l were then updated using the 1990 census data

• and projected to the year 2010 by decades. The Revision 8 2.1-6 l

influence of transient populations, and locations of population centers, public facilities and institutions in the vicinity of the Trojan Nuclear Plant are also defined.

2.1.3.1' Population Within 10 Miles The 1970 population distribution within 10 miles of the site is based on the 1970 census,

-1 7-1/2-min.,1970 photo revised, U.S. Geological Survey (USGS) maps, and the most recent {

f issues of the 15-min. USGS maps. The procedure used was to divide the rural population in each census division by the number of rural structures in each division to obtain the average number of

. people per structure. The population in each sector was 1

then determined by the number of rural structures within the sector added to the town and city populations. All incorporated areas are shown as well as unincorporated areas with 1000 or more population.

. The population projection for 1980, within 10 miles of the site were made by the Economic Development and Research Department of PGE. The forecast was made county by county based on historical population data from 1900 through 1970. The projections from strictly historical data were modified by reviewing other projections", and the judgment of the analyst  !

concerning economic opportunities, demographic factors, land use patterns and the physical ability of the area to absorb additional population. The forecast is consistent with the U.S.  ;

Bureau of the Census, Series D.

Population gro'wth factors s.ere obtained for each county and for each year considered. The factors were then applied to tt '970 population for parts of one or more counties falling within a sector to obtain an estimatedi ~,ected population within that sector.

A review of basic demographic data from the 1980 census for the area within 10 miles of Trojan was undertaken to determine if updating the 1980 population projections was necessary. This review concluded that the 1980 projections are conservative.

2.1-7 Revision 8

The 1990 population distribution within 10 miles of the site is based on the previous efforts updated to reflect 1990 census values"). Specific place populations were located within the appropriate sectors. Rural population groups were distributed on the basis of the density of roads within each sector.

The population projections for 2000 and 2010 within 10 miles of the site were made using county growth projections based upon three census data points: 1970,1980, and 1990. Individual growth projections were developed for Cowlitz and Columbia Counties. Based upon these historical factors, population growth within 10 miles of the site is about 5 percent per decade. This moderate growth does not take into account the recent changes in population changes due to the changing economics of the area.

2.1.3.2 Population Between 10 and 50 Miles The 1970 population distribution between 10 and 50 miles of the site is based on the 1970 census and the most recent issues of the 15-min. USGS maps. Estimates of the population distribution were made using census division tabulations and house densities.in each sector. All incorporated areas, and unincorporated areas with greater than 5000 population, are shown.

The population projection for 1980 between 10 and'50 miles of the site was made in a manner similar to tha't described in Section 2.1.3.1. ,

The 1990 population distribution between 10 and 50 miles of the site is based upon current county-by-county population estimates"). Specific place populations were located within the appropriate sectors. Rural population groups were distributed on the basis of the density of roads within each sector.

The populations for 2000 and 2010 were made in a manner similar to that described in Section 2.1.3.1. It should be noted that the projections for the Portland area are likely to Revision 8 2.1-8 i

overstate the actual figures that will result in those years. That is due to the fact that the A

V projections are based on past growth and the large growth rates of the past 20 years are unlikely ,

to be maintained.

i 2.1.3.3 Transient Population 4

The major source of population influx and shiftr occurring within 50 miles of the site are the

. result of recreational activities that peak during the summer months. Information was not available as to the number of people entering the area from outside the 50-mile i i

radius (influx) or the number corresponding to shifts within the 50-mile radius. Most of the park

~

facilities are further from the site than the two major population centers of Vancouver, Washington and Portland, Oregon; therefore, the seasonal population shift within the 50-mile radius would tend to be away from the site"*.

There are no State or Federal parks or campgrounds within the low-population zone"*, which tends to limit to reasonably small numbers the influx of fishermen and campers. There is a limited influx of people into the low-population zone ~during the spring and summer months when the river conditions are conducive to good fishing. This influx is primarily on the Columbia and Kalama Rivers and consists of pleasure boaters, boat fishermen and bank fishermen"". From the observed number of boat and bank fishermen on the Columbia River between Prescott (River Mile 72) and the Longview Bridge (River Mile 66) from March to September 1971 it is estimated l that an average of 121 fishermen per day are present inside the low-population zone from March I through September (74 on the bank and 47 in boats). During September, the month of highest i

utilization, an average of approximately 231 fishermen per day are present on the Columbia River l inside the low-population zone.

There are no known agricultural activities within the low-population zone that might cause seasonal population changes.

O 2.1-9 Revision 8

r Population shifts during the working day will tend to be out cf the low-population zone and l

towards the cities and towns surrounding the site. l 2.1.3.4 Low-Population Zone l

The low-population zone consists of the area outside of the exclusion area but within a 2-1/2-mile 1

radius of the reactor vessel. This distance was selected since it meets the requirements of ll 10 CFR 100.

The 1970 population distribution within the low-population zone was determined in the same way as explained in Section 2.1.3.1. .

The population projection for 1980 within the low-population zone was made as described in Section 2.1.3.1.

The population projections for 2000 and 2010 were r. de in the manner described in Section 2.1.3.1.

2.1.3.5 Population Center The nearest population center consists of the cities of Kelso and Longview, Washington. These two adjacent cities form a community with a combined population in 1990 of 43,319, with a near-side distance 6 miles due north of the site.

2.1.4 USES OF ADJACENT LANDS AND WATERS The Trojan Nuclear site lies in a heavily timbered area, characterized by rough terrain and suited primarily to logging and other forestry operations. One major population center lies within this radius, as do several smaller cities; most heavy industry in the smaller cities is related to forest Revision 8 2.1-10

. product processing or agriculture". Well over half of the land is suitable for commercial forestry or grr. zing, with about 20 percent suitable for farming. Less than 10 percent of the land area is unsuitable for any agricultural pursuit, and a fraction of 1 percent is devoted to urban or f

incorporated areas.

Lands adjacent to the site lie within Columbia County, Oregon, in which the site is located, and Cowlitz County, Washington, across the Columbia River. All of the area within a radius greater than 10 miles from the site lies within these two counties. Both have agriculturally based economies, with land use in the vicinity of the Trojan site primarily agricultural. Logs, hay and other feed are the predominant crops. Salient agricultural data for these two counties is indicative of small, generalized family farming, with heavy emphasis on gr'azing and farm animals. Only 41 of 934, less than 5 percent, of the farms in the counties are larger than 100 acres, while more than one-third have fewer than 20 acres. There are no major milk-producing centers on lands adjacent to the Trojan site, the major milksheds being 50 or more miles distant".

Distances were measured from the center of the containment to the closest site boundary, residence, garden, meat animal, milk cow and milk goat within a' 5-mile radius of the Plant for

. each of the 16 directional sectors. The field survey from which these distances were determined was performed in September 1980.

Other than agriculture, the industrial base of the area around the Trojan Nuclear Plant is centered

- in forest products and primary metals, and most of the industrial activity is on the other side of the Colurabia River. Of the eight manufacturing firms in the Longview-Kelso area employing more than 200 persons, five are forest products processors producing lumber, plywood, pulp, paper and paper products, and related wood and wood pulp productsa4 .23.2e. The other large manufacturing firm is an aluminum smelter with an annual capacity of 204,000 metric tons". A small steel furnace smelter and pleasure boat manufacturer are the only other major manufacturers in this area. Near Kalama, upstream of the site, are grain elevators, chemical O 2.1-11 Revision 8

r plants and a few small mills. No synergistic effects on the environs are observed as a result of the Trojan Nuclear Plant dnd these industries.

There is relatively little recreational land use within the immediate area of the site. There are no State or Federal parks nearby, nor are there any natural or man-made attractions such as mountains, dams or lakesm .29). A 3-mile portion of U.S. Highway 30 has been designated as a scenic area by the Oregon State Scenic Area Board"). Pleasure boat launchings are located in Rainier, about 4-1/2 miles due north of the site, in Goble,2 miles SSE of the site, and in Prescott,1-1/2 miles NW of the site 00 Recreational vehicle overnight parking is available in Goble, and Prescott Beach, located 1 mile NNW of the site, is used for camping and fishing.

River access is also available on the Washington shore, directly opposite the Plant.

Te lower Columbia is a river well suited to recreational fishing and boating, almost all of which occurs in the 7 months from March to September. The Oregon State Department of Game estimates approximately 30,000 angler days of river use in the stretch from Prescott to the Longview Bridge, approximately 6 miles downstream of the site o2> . Only about one-third of this usage is attributable to boat anglers. The much longer river stretch from Corbett, about 60 miles upstream, past the site to Prescott, has an estimated usage of about 160,000 angler days. Thus, average usage is about 250 to 500 angler days per mile of river per year.

. Commercial fishing in the Columbia is regulated by both Oregon and Washington. About 270 miles of the Columbia River and tributaries are open to commercial fishing, with Bonneville Dam be.ing the approximate midpoint. The fishery upriver of Bonneville is reserved for Indians only, while downstream i.s open to commercial fishing license holders as well.

The Columbia River is a major navigable channel. Approximately 2300 seagoing cargo vessels pass the Trojan site annually, carrying wheat and logs outbound and manufactured iron goods, ores and petroleum inbound'". Major port facilities are at Portland, Oregon and Longview, Washington. Commercial river traffic is discussed in detail in Section 2.2.1.

Revision 8 2.1-12

I 2.2 NEARBY INDUSTRIAL. TRANSPORTATION AND MIT ITARY FACII ITIES Potential accidents as a result of extraneous activities in the vicinity of the Trojan Nuclear Plant -

havf. been studied to determine their effect on the safety of the Plant. This section outlines the activities of the nearby industrial facilities, transportation arterials and military installations and their potential effects on Plant safety. The risk to the operation of the Plant resulting from these activities is shown to be minimal.

2.2.1 LOCATIONS AND ROUTES Most of the local commerce is related to forest products and is centered in three areas:

Longview, Washington, approximately 6-miles northeast of the site; Rainier, Oregon, approximately 4-1/2-miles northwest of the site; and Kalama, Washington, approximately 2- l 1/2-miles southeast of the site.

Due to the emphasis on forest products, industrial development in the area is heavily oriented to river transportation. An aluminum plant in Longview, chemical plants and grafn elevators in Kalama, and a fertilizer plant in Columbia City are the only large industries not related to the timber or paper industry. There are also several small quarry sites and gravel pits in the area, the closest being in Goble, approximately 1-mile southwest of the Plant.

Transportation routes consist of two major highways, two railroads, the Columbia River and an airport and airways. U .S. Highway 30 runs nonh-south adjacent to the site boundary approximately 2400 feet from the Fuel Building , and is a light-duty, two-lane highway connecting Portland on the south to Astoria, at the mouth of the Columbia River. Interstate 5 (I-5) is part of the West Coast north-south interstate system planned to extend from Mexico to Canada. I-5 in this area is across the Columbia River in Washington approximately 1-1/4-miles east of the Fuel Building at its nearest point. The Portland & Western Railroad right-of-way l passes through the site, approximately 600 feet from the containment. The main line railroad '

O V

2.2-1 Revision 8 m_

track between Portland and Seattle is located across the Columbia River in Washington, approximately 5000 feet from the site.

The Cclumb,ia River, one of the largest rivers in the nation, serves as the deep-sea access channel to the important ports of Poniand, Oregon and Vancouver, Washington. A 40-foot channel is maintained for deep-draft ocean vessels as far upriver as Portland. The center line of the 600-foot wide ship channel is approximately 1400 feet from the site. Upstream from Portland and Vancouver, a 17-foot channel is maintained for barge traffic, extending to Pasco, Washington and a distance into the Snake River m . Locks are provided at each of the dams on the river *.

The Kelso-Longview Airport is 5.3-miles nonh of the site and has a 4395-foot paved runway oriented northwest-southeast. .

The Portland International Airport is located 33 statute miles south of the site, and is the only major airport within a 60-mile radius of the site.

There are no major military bases in the vicinity of the Trojan Nuclear Plant. The nearest O

military facilities are Reserve Headquaners for the various branches in Ponland and Vancouver (30-40 miles south of the site), Beaver Army Terminal (20 miles northwest of the site) and Coast Guard and Naval facilities in Portland, Longview and at the mouth of the Columbia River *.

l

[ U.S. Highway 30 provides highway access to the Trojan site and serves as the traffic arterial l

l between Portland and the communities on the Oregon bank of the Columbia River, carrying an average of 5300 vehicles per day *. The highway runs through the communities of Scappoose, Warren, St. Helens, Columbia City, Deer Island and Goble, south of the site; and Rainier, Clatskanie, Westport and Astoria north and west of the site. A bridge at Rainier connects U.S.

Highway 30 with Longview, Washington, and a bridge at Astoria, the western terminus of the l highway, connects to Megler, Washington.

Revision 8 2.2-2 O

________.__________A

U.S. Highway 26 provides a shorter Portland-to-Astoria route; thus it carries the bulk of traffic A

V between the two, leaving U.S. Highway 30 to carry local passenger traffic, log trucks, tourists, farm vehicles and truck deliveries to the river communities. There is some shipment of petroleum products via U.S. Highway 30. Gasoline, diesel and heating oils in tank tmcks are regularly delivered to towns beyond Trojan from suppliers in Portland and St. Helens.

Interstate 5 is the primary. north-south traffic route between Portland and the Puget Sound area (Seattle, Tacoma, Olympia). An average of approximately 46,000 vehicles per day. Of this total, approximately 20 percent is made up of truck combinations and the remaining 80 percent is ,

i passenger traffic *. It is estimated that about one-tenth of the tmck traffic could be carrying  !

~

flammable or hazardous material, of which petroleum products would make up the majority.

An average of two freight trains per day pass through the site on the Portland & Western l l

Railroad right-of-way, carrying general commodities, with an annual gross tonnage of 6 million tons *. Lumber and forest products make up the bulk of the shipping most of the year. During Y the peak fishing season, some canned and frozen seafood is carried by rail from the Astoria canneries. An average of about 200 shipments per year with 2-3 cars per shipment of chlorine l and caustics are shipped to the James River Corporation in Wauna, Oregon, on the lower river via the Portland & Western line. Other chemicals shipped include preservatives, fertilizer, resins l and paints and a small amount of petroleum and propane.

Three railroads use the tracks on the Washington side of the river: the Burlington Northern,

' AMTRAK, and the Union Pacific railroads. Thirty-five to forty freight trains and six passenger trains pass the site per day on these tracks m . The freight carried varies widely with large quantities of wood products, aluminum, paper products, grains, agricultural products and

~ foodstuffs making up the bulk. Chemicals shipped include large quantities of fertilizers, phenols, caustics, propane and various resins, acids, paints and lumber treatments.

O 2.2-3 Revision 8 a

Sharply rising ground to the west and similar high ground across the river to the east provide l natural barriers for the site. The highest permanent structure at the site is the cooling tower, -

rising 499 feet above ground level. The cooling tower is marked in accordance with Federal Aviation Administration regulations, with flashing beacon lights at the top and midheight, and obstmetion lights at the quarter height. There are four lights at each level, equally spaced around the perimeter. In addition, high-intensity flashing strobe lights are equally spaced around the top of the tower.

2.

2.2 DESCRIPTION

S 2.2.2.1 Description of Products and Materials Products and byproducts of the timber industry in the area range from unfinished timber to finished construction lumber, cabinetry, plywood and veneer. Some hardwood products are made in Longview on a small-scale operation, while paper and wood fiber products make up a large percentage of the production of the area. Some chemical use and storage is associated with these industries. Chemicals include resins used in plywood, veneer and chipboard production, acids used in paper and pulp production, and lumber pressure treatments and finish coatings (stains and varnishes). Chemicals are stored either in tank cars on sidings, or in storage tanks connected to the industry involved'8) .

i The aluminum plant in Longview consists of two facilities: an aluminum reduction facility operated by Reynolds Metal Company which produces raw metal in the form of ingots, billet bars, etc., and a cable-production facility operated by Cablec Utility Cable Company which produces electrical conductors and aluminum wire. The use of chemicals at these plants correspond to that of any aluminum plant; namely coke, pitch, chlorine and liquified nitrogen.

Chemical storage facilities at the Plant consist of stockpiles, tanks and rail tankers and transportation is by rail tank cars (8') .

1 Revision 8 2.2-4 O ,

t. i a

Kalama Chemical, Inc., produces greater than 50-million pounds per year of phenols with some 7 secondary production of benzoates. The facility receives its raw material, toluene from tankers and stores it in an 80,000-barrel tank. The finished product is shipped by rail tank car *.

Hoechst Celanese Corporation, Inc., is located approximately 3-miles southeast of the' Plant in Kalama, Washington and produces a bleaching agent used in the pulp and paper industry. The facility receives sulfur diqxide by rail tank car and has a storage capacity for this chemical of ,

500,000 pounds. i All Pure Chemical Company is located approximately 2-miles southeast of the Plant in Kalama, Washington. The company produces a number of products including sodium hypochlorite, l household ammonia, and water treatment chemicals. It is involved in tbe repackaging and I

distribution of chlorine gas. The chlorine gas is received in 90-ton rail tank cars and is l

- repackaged into 1-ton cylinders. The 90-ton rail tank car is the maximum storage capacity for the chlorine gas at the facility.

~3

(

2.2.2.2 Pipelines A natural gas main extending to Wauna, Oregon, downriver of the site mas along the hillside west of the site, approximately 1-1/2 miles from the site . The main is a 16 inch,3-million foot / hour line, buried a minimum of 3 feet" ). In addition, there is an odorizer station on the line at Goble,2-miles south of the site, a river crossing at Deer Island,4-1/2-miles south of the site and a river crossing at Rainier,5-miles north of the site.

2.2.2.3 Waterways The Columbia River, one of the largest rivers in the nation, serves as the deep-sea access channel to the important ports of Portland, Oregon and Vancouver, Washington. A 40-foot channel is 2.2-5 Revision 8

maintained for deep-draft ocean vessels as far upriver as Portland. The center line of the 600-foot wide ship channel is approximately 1400 feet from the Fuel Building.

About 2300 oceangoing ships a year pass the site on the Columbia River. The major portion of l the cargo exported is wheat and logs. Inbound cargo consists of miscellaneous goods such as petroleum, iron and steel products and ores"". Portland is one of the largest ports in terms of tonnage on the Pacific Coast and thus it maintains a large number of supporting facilitie.s.

Longview, Washington, downstream of the Plant also has facilities for oceangoing ships. The l Port of Longview maintains facilities for unloading and storage of ship cargo. Significant facilities are a bulk loader with storage for 14,000 metric tons of tale; storage tanks with capacity for 40,000 tons of calcinated coke; a grain elevator, currently not in use, with a capacity of 7.8 million bushels; and log storage yards. Among the commodities routinely stored at the port are pencil pitch (or coaltar pitch), a ammonia sulfate, and potash. Additionally, at the port site, Wilcox and Flegel operate a petroleum receiving and bunkering facility which has 14 storage tanks with a total capacity of 26,190 barrels of storageo2>, h 2.2.2.4 Airoorts The Kelso-Longview Airport is 5.3-miles north of the site and has a 4395-foot paved runway oriented northwest-southeast. The airport is not a scheduled airline stop but is the base for approximately 75 single and twin-engine, private and corporate aircraft. The airport handles about 100 takeoffs and landings per day. The largest planes using the field are a DeHavilland 8 corporate plane, a Siddely Hawker, a Cessna Citation, and a Falcon Jet"'). The Portland International Airport is located 33 statute miles south of the site, and is the only major airport within a 60-mile radius of the site. Portland inbound and outbound air traffic is controlled for a distance of 30 miles from the airport by Portland Air Traffic Control. Area-wide inflight traffic control is regulated by Seattle Air Traffic Control"*

Revision 8 2.2-6 4

l 2.2.3 EVALUATION OF POTENTIAL ACCIDENTS b i This section provides an evaluation of the capability of the Trojan Nuclear Plant to safely withstand the effects of an accident at, or as a result of the presence of, industrial, transportation and military installations or operations within 5 miles of the site. Pote' ntial accidents considered include impacts to and blockage of the cooling water intake structure; explosions of chemicals, flammable (including natural) gases or munitions; industrial and forest fires; accidental releases of toxic gases; accidental releases of corrosive liquids or oil into the Columbia River upstream of the cooling water intake structure; and collapse of the cooling tower.

2.2.3.1 Frnlosions Shipments of commercial cargo past the Plant create the possibility of nearby explosions.

Safety-related portions of the Plant are protected from such explosions.

Explosions unrelated to transportation are not considered significant. The quarry operations south of the site are located in the hills west of the Co'lumbia River. Presently, there is no storage of explosives at the operating quarry which is 2 miles from the site. The quarry is not a large operation and only a limited amount of explosives are used. Because of the distance from the Plant and the protection afforded by the hillside and ridge between the quarry and the Plant, the quarry operation does not present a hazard to the safety of the Plant. A natural gas main extending to Wauna, Oregon, downriver of the site runs along the hillside west of the site, approximately 1-1/2 miles from the site. The main is a 16 inch, 3-million foot / hour line, buried a minimum of 3 feet"*. In addition, there is an odorizer station on the line at Goble,2 miles south of the site, a river crossing at Deer Island,4-1/2-miles south of the site and a river crossing at Rainier,5-miles north of the site. The operation of these lines will not present a hazard to Trojan from explosion because of the relatively low explosive capacity of the gas and the distance from the Plant proper.

,e

Even though there is some evidence that transportation accident rates involving hazardous cargo are below the overall average because of precautionary measures taken with such shipments"", no credit is taken for this factor.

On the other hand, it may be assumed that the rate of accidents caused by the nature of the hazardous material itself (explosion of cargo not caused by traffic accident) is comparatively low and may be neglected. However, it is realized that the transportation accidents down to a low degree of severity must be considered here as the explosion may be triggered by a minor cause (shock, spark, heat).

~

The effect of a transportation accident involving explosive cargo is based on the assumption that an explosion does occur.

The major distance effects of a surface explosion are:

(1) Atmospheric effects (shock wave and momentum effects).

(2) Ground effects (direct and air-induced ground shock).

(3) Other effects (missiles, waves, noxious fumes or incendiary effects).

Studies of these effects have led to the conclusion that the atmospheric shock wave is the severest effect (most limiting) with regard to safety-related structures of power plants"6 * .

The value of the limiting (maximum permissible) overpressure generated by the atmospheric shock from the explosion is 2.2 psi (p, = 2.2 psi). This limit is chosen in recognition of the fact that the actual overpressure experienced by parts of the structure may be up to twice this value (4.4 psi) due to reflected wave effects"U.

Revision 8 2.2-8

It is noted that test structures generally similar to those employed for safety-related structures at Trojan have' demonstrated that overpressures of 5 psi can be endured without any significant damage to these structures"*. Thus, it appears that the 2.2 psi overpressure used in this analysis incorporates substantial conservatism. Furthermore, it is noted that safety-related structures are designed to withstand pressure differentials of at least 3 psi due to tornado effects.

A missile caused by an explosion at a distance greater than that at which 2.2 psi would occur would not be as severe as the assumed tornado missiles. If the range were less than that corresponding to 2.2 psi, the blast damage would be more critical than the missile damage.

Missiles from the cooling tower failure are considered comparable to the tornado missiles. If the probabilities of the explosion occurring within the range corresponding to 2.2 psi are shown to be remote, the effects of the explosion can be neglected.

The method of analysis employed subsequently is common to all types of transportation. This is due to the fact that the means of transportation considered here (train, barge) are confined to a rather narrow track (rail, channel) which leads past the Plant nearly as a straight line. Trains and barges are assumed to be moving along these lines at normal speed and are therefore subject to

- accident rates per mile of track (taken to be constant along the track). The pertinent accident rates are discussed below.

{J \

2.2-9 Revision 8 l

= - - _ - - - .

l

An approximate relation for the peak (positive) overpressure of the shock wave from an explosion can be written as: I p = a x g* (2.2-1) where i p = peak overpressure (psi) f g = scaled distance from cer f explosion (ft/lb"$)

a,.b = constants.

I A good approximation in the range of pressure 0.1 < p < 10 psi is obtained by setting:

a = 200 O l l

The scaled distance is defined as: I g = _P_ (ft/lb"') (2.2-2)

W ") {

I where P = distance from center of explosion (ft)

W = weight of explosive, TNT equivalent (Ib).

Revision 8 2.2-10 i

l (3 From these relations we obtain:

G p = 200 x W"'6 P (psi) (2.2-3) or else: i l

I P = 45 x W"" x pez (ft) I (2.2-4) and:

5 2 W = 1.1 x 10 x p . ' x P' (lb) (2.2-5)

- For the limiting overpressufe of p, = 2.2 psi, the corresponding scaled distance becomes:

. g, = 25.8 (ft/lb"3)

The limiting distance corresponding to 2.2 psi overpressure is given by:

De = 25.8 x W"" (ft) (2.2-6)

The preceding equation for W may be used to determine the minimum weight of explosives which l

may have a disabling effect on the Plant when exploded anywhere on each one of the respective lines of traffic considered here. These minimum weights are obtained by using the nearest approach distance for the value of P and the respective limiting overpressure.

A probabilistic approach must be applied when assessing hazards from transportation accidents involving explosives on the Portland & Western (formerly Burlington Northern) Railroad and on l the Columbia River. For the purpose of this analysis it is assumed that explosive cargo may be

{s 2.2-11 Revision 8 l

4

f classified as belonging to one of several categories, each category (i) characterized by a representative weight of the cargo (W)i and the annual number of shipments (f). i The annual probability of a transportation accident involving explosive cargo with a disabling effect on the Trojan Plant is determined below:

P = Pr X P, X E if Li (year) (2.2-7) where P = annual accident probability ,

Pr = accident rate per mile (rail, river) p, = probability that accident will lead to explosion of cargo f i = annual number of shipments (weight category i)

L i = length of track (miles) over which cargo (of weight category i) is within critical distance P of Plant.

(*

Revision 8 2.2-12 l

l

For a straight track, as is generally the case near Trojan, Li can be determined from:

O -

'L,'=

5280 X}D*-dz (2.2-8) where

. d = closest approach of track to Plant (feet). '

When considering weight categories much in excess of the miniinum weights, it is a conservative and good approximation to use instead of the preceding equation, the simplified relation:

i l

Li = _2 x P (miles) i (2.2-9) )

5280

)

O V Inserting the relation obtained earlier for P, we arrive at:

1

'L i = 0.0098 W""(miles) i (2.2-10) l The corresponding equation for the annual probability is:

P = Pr X P. x 0.0098 iS f, x W"" i (yr) (2.2-11)

The accident rate per mile of train travel is found to be 0.9 x 104 (miles- )"S. This value is based on accident statistics, considering experience nationwide as well as Burlington Northern Railroad only, which has an accident record somewhat better than (below) the national average. Data for 2.2-13 Revision 8 J'

,' i

1969-1972 were considered and show a rather constant accident rate. Of the three major accident categories (train , train service , nontrain-accidents) only the first was considered to be pertinent. .

The value quoted above is believed to be highly conservative for the following reasons:

(1) Most of the accidents included in this figure are of very minor degree (1/3 of the i accidents incur damage below $10,000)

(2) Many of the accidents included in this figure occur in rail yards (1/3 of the accidents)

(3) On the average, only 1/7 of the cars in a train are affected by the accident.

In the face of this evidence, it is believed that a value of 4

pg = 0.2 x 10 (miles )

O is a more realistic value for the pertinent accident rate per mile of train travel yet holding a degree of conservatism.

The probability p, for a rail accident involving explosive cargo resulting in an explosion of this

- cargo is conservatively estimated as:

p, = 0.1 Based on these figures one would expect 16 explosions per year from rail accidents in the United States assuming that 1 out of 10 trains carries explosive cargo. Information received from the Bureau of Explosives indicates that there were 44 explosions in railroad accidents in a recent 13-year period. The average of 3.4 explosions in rail accidents per year derived from this information points to the degree of conservatism underlying the assumptions made above. ,

Revision 8 2.2-14 h

> l

Previous studies have been based on barge accident rc.es per mile of 1.8 x 104 (miles'8)

O <aer r ce i>> e s.6s x io <-iies ') <a r r "ce 16). it e 8ee >t- t a di tde cei -di- -

River Pilots Association"" that there have been some 15-20 mishaps on the lower Columbia River during the last 10 years, most of them of minor degree. A distance of about 100 miles is involved with an estimated average of 3000-5000 shipments per year. The estimated accident 4

rate per mile based on these figures is about 4 x 10 (miles), which is in agreement with the above values. Thus, we use for barge traffic:

4 Pr = 4 x 10 (miles ')

The probability of a barge accident involving explosive cargo resulting in an actual explosion is believed to be more remote than in the case of a train accident. Pre.vious studies have been based conservatively on a value of 1/20 for the - 'sbility"*. This value shall be used in the further analysis:

p, = 0.05 O ~ The analysis above has shown that only explosive cargo in excess of about 1 ton needs to be of concern for rail shipments. Accurate statistics of cargo shipped on the rail under concern are not readily available. However, conservative estimates by industries serviced by this rail line show tb 1 inert: is an average of about one shipment per month of one (or more) tank car of propane or

, equivalent 22) . On this basis, we have:

f i = 12 (year ~ )

Wi = 55 (tons) = 110,000 x 126 821

= 70,000 (pounds of TNT equivalent)

According to information received from Burlington Northern"3), there have not been any shipments of high explosives on this rail line in the past, and no routine shipments of cargo are 2.2-15 Revision 8

1 1

expected for the future as explosives (needed for logging in the vicinity) are shipped by truck (U.S. Highway 30, I-5).

It is noted that oil derivatives such as propane may not be comparable to high explosives as they do not contain sufficient oxygen for detonation. Thus, the explosion of a propane tank will have a milder effect than that of an equivalent amount of TNT under normal circumstances.

There are similarly no accurate data available for the shipment of explosives on the Columbia River in recent years. According to the Columbia River Pilots Association", there has been one shipment of high explosives on the affected stretch of river (Astoria to Portland) in recent years.

The Umatilla Army Depot, located 210 miles upstream of the site, is the only facility on the Columbia River capable of handling Class A explosives *. The depot is currently served by rail and truck transportation only". Even if a change in depot operations occurred, only the largest practical shipment of explosives, according to local barge companies, is equivalent to 2500-3000 tor.s of TNT. A cargo of 3000 tons of TNT exploding in midstream opposite the intake structure would impose approximately 15-psi peak pressure on the sup:rstrue:ure, which the 2-foot-thick concrete walls of the structure could withstand safely. An explosion of a chemical or gas tanker would have a les'ser effect on Spent Fuel Pool (SFP) activities. Thus, we assume for the purpose of obtaining a conservative e, stimate:

fi = 1 (year)

6 Wi = 3000 (tons) = 6 x 10 (lb)

Based on the methods and assumptions outlined above, the following estimates of the annual probability of a disabling accident can be made as follows:

P = pr X P. x 0.0098 xi f W i " (2.2-12)

Revision 8 2.2-16

r 4

Railroad: P = 0.2 x 10 x 0.1 x 0.0098 x 12 x 41.2

\

< 1.0 x 104 (year) '

Barge: 4 P = 4 x 10 x 0.05 x 0.0098 x 1 x 182

< 1.0 x 10# (year)

It is estimated in a highly conservative manner that the annual probability of a transportation accident near the Plant involving an explosion with a potential for a disabling accident is less than one in a million. It must be pointed out that this probability calculation assuned reactor plant operations. With the permanent Plant defueling caly those systems associated with the storage of irradiated fuel and radioactive waste are considered in DBA. This has resulted in a significant reduction in the number of systems / components potentially affected by an explosion. The consequences of a DBA have also been significantly reduced such thst snalyzed doses remain below 10 CFR 100.

I The conservatism incorporated in every step of the analysis gives a high degree of confidence that ,

the probability estimates given above constitute upper limits and that more realistic probability values are significantly lower than those obtained here.

2.2.3.2 Toxic Chemients .

2.2.3.2.1 Onsite Storage l Limited amounts of toxic chemicals are stored on site. Hydrazine may be stored at Elevation 45 feet of the Turbine Building, and also in drums in the onsite warehouse. This chemical has a low vapor pressure, and could only constitute a hazard in the event of a fire. Therefore, no specific protective actions are needed against this substance. Sodium hypochlorite is also stored onsite. The potential quantities of chlorine from sodium hypochlorite are so small that this 1

release would have no impact on the control room. '

j 2.2-17 Revision 8

2.2.3.2.2 Offsite Storage and Transportation i

Four storage locations for toxic materials exist wititin 5 miles of the site: Longview Fibre Company (5 miles north-northeast), Kalama Chemical Company (2 miles southeast), Hoechst Celanese Corporation (3 miles southeast) and All Pure Chemical Company (2 miles southeast).

The principal transportation routes within 5 miles of the Plant are:

Columbia River (at the closest point to 400 meters the control room) -

U.S. Highway 30 800 meters Interstate 5 2000 meters l Portland & Western Railroad west of Plant 150 meters Railroad east of Columbia River 1800 meters l

Since the railroads are closer to the Plant than the highways and since railroad shipments of toxic chemicals are larger than highway shipments, it was concluded that the railroads are the critical land transportation routes from the standpoint of control room habitability.

l It was determined that the only toxic chemical shipped by the Portland & Western (formerly j l Burlington Northern) Railroad on the west side of the Plant with sufficient frequency to require analysis" is chlorine (about 200 shipments per year) *.

I Toxic chemicals are also shipped via the railroad track eau of the Columbia River by the Burlington Northern and Union Pacific Railroads .27 a 33>. Chlorine, sulfur dioxide, sulfuric acid, methanol and ammonia are shipped with a frequency and in weights sufficient to require analysis". .

Revision 8 2.2-18

1 l

I

2.6 REFERENCES

REFERENCES FOR SECTION 2.1

1. Telecon No. WRN-T15-92, W. R. Nehrenz to D. Hefty at Merchant's Exchange.

2. Number ofInhabitants. Oregon, PC(1)-A39 Oreg. U.S. Department of Conunerce l (July 1971). l l

3. Number of Inhabitants. Washington, PC(1)-A49 Wasi.. U.S. Department of Commerce l (June 1971). l ,

I i

4. 1990 Census of Population. U.S. Census Bureau. l l

5. Population and Hon =ehold Trends. 1960-1980. . Pacific Nodhwest Bell (April 1%5). l l l

6. Population and Household Trends in Washington. Oregon and Northern Idahn. 1960-1980. l Pacific Northwest Bell (March 1%7). l l

7. Population and Household Trends in Washington. Oregon and Northern Idaho. 1970-1985. l Pacific Northwest Bell (January 1972). l O 8 . Forecast prepared for the Oregon State Highway Department by the Oregon Division of l I I

Planning and Development,1966.

9. Based on Population Bulletin P-10', Oregon State Board of Census released April 1964.

10. Population Prosoects. Portland metropolitan Planning Commission (1960). l

11. Mid Willamette Valley Council of Governments, unpublished data (December 1967). l

12. Rivergate and the North Portland Peninsula. Daniel, Mann, Johnson, and Mendenshall l (1967). Based on rivergate Industrial Demand Study, Battelle Memorial Institute (1%5).

13. Projections prepared by Mid Willamette Valley Council of Governments (March 1972).  !

)

14. Projections prepared by Boatweight Engineering, Salem, Oregon, for Columbia County .

l (1970), before data from the 1970 U.S. Census of Population was available.

15. Economic Profile with Projections to 1990. Portland-Vancouver SMSA. Columbia Region f Association of Governments (August 1%8).

A V 2.6-1 Revison 8

l 16. Oregon State Park Visitor Survey - 1969. Oregon State Highway Division, Parks and

~

Recreation Section. ,

l 17. Washington State Outdoor Recreation Guide. Washington State Parks and Recreation Commission.

18. Washington State Parks and Recreation Commission Attendance Figures for June 1971 through May 1972.

l 19. Oregon State Parks and Waysides. Oregon State Highway Division State Parks and Recreation Section (1972).

20. Oregon State Highway Division, State Parks Section, Attendance for the years 1966 through 1971.

l 21. Elc.asure Boating in Oregon. Oregon State Marine Board (July 1,1972).

22. Derived from U.S. Department of Commerce, Bureau of Census',1987 United States Census of Agriculture - County Data, Oregon Columbia County, pp 167-17'7; County Report for Columbia County (December 9,1991); Washington-Cowlitz County, pp 65-72 (September 1961).

l 23. State Milk Surveillance Activities. Radiological Health Data and Reports,9:637 (1968).

l 24, 1967 Census of Manufactures. Area Series. Washington, Mc67(3)-48. U.S. Department of Commerce, Bureau of the Census (November 1970).

l 25. Washington State Standard Community Industrial Survey. Kelso-Longview. Washingto:

(January 1972) l 26. Community Economic Profile for Longview. Cowlitz County. Wa. Longview Chamber of l

Commerce (July 1991).

l l 27. The Longview Daily News. Special Advertising Section, Longview Chamber of Commerce >

(September 16, 1991).

l 28. Atlas of the Pacific Northwest,4th Edition. Richard M. Highsmith and Jon M. Leverenz, Ed. Oregon State University Press, Corvallis, p 155-159 (1968).

l 29. Planning for Parks - Columbia County. Oregon. Bureau of Governmental Research and Service, University of Oregon. A Report for the Columbia County Planning and Park Commissions (June 1968). ,

1 i

Revision 8 2.6-2

)

l l

30. Ertlininary Road Plan Columbia County. Oregon. Bureau of Governmental Research an?

n. Servicc (March 1967).

l t]

31. Gary Fritz, Caretaker-Prescott Beach Park, Telecon (July 1992).

32.

John Haxton, Oregon State Game Commission, Personal Communication (July 1972).

33.

Trojan Nuclear Plant Final Safety Annivsis Reoort, Amendment 19 (December 1992).

l REFERENCES FOR SECTION 2.2 1.

Telecon No. WRN-T19-92, W. R. Nehrenz to S. Perkins, U.S. Army Corps of Engineers.

2. Corps of Engineers, Portland. -

3. U.S. Department of Defense (1972).

4. Telecon No. WRN-T20-92, W. R. Nehrenz to H. Nale, Oregon Department of Transportation, Planning Section.

5. Telecon No. WRN-T21-92, W. R. Nehrenz to J. Sorrell, Washington Department of Truisportation, District 4 Traffic Office.

6. Ken Pellens, Roadmaster, Burlington Northern Railroad, Telecon (June 1992).

7. John Myers, Train Master Vancouver Office, Burlington Northern Railroad (June 1992).

8. Survey Conducted by PGE in Trojan Area (1972).

9. Telecon No. WRN-T44-92, W. R. Nehrenz to W. Maines, Cablec Utility Cable Company.

10. Northwest Natural Gas Company. (1972).

11. Telecon No. WRN-T15-92, W. R. Nehrenz to D. Hefty, Merchant's Exchange.

12. Telecon No. WRN-T38-92, W. R. Nehrenz to K. Barney, Port of Longview.

13. Harry O'Neil, Kelso-Longview Airport Manager, Telecon (June 1992).

14. Federal Aviation Administration (1972).

TT 2.6-3 Revison 8

1 1

l l 15. Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants Directorate of Regulatory Standards, U.S. Atomic Energy Commission, d (December 1972).

l 16. Brunswick Steam Electric Plant. Units 1 and 2 PSAR, Amendment 1, Supplement 1. USAEC Docket 50-324-3, Carolina Power and Light Company, Raleigh, North Carolina (September 1968).

l 17. Sequoyah Nuclear Plant PSAR, Amendment 4. Docket 50-327/328, Tennessee Valley Authority, Knoxville, Tennessee (May 1969).

l 18. Principles and Practices for Design of Hardened Stmetures. M. N. Newmark and J. D. Haltiwanger. Air Force Special Weapons Center Report AFSWC-TDR-62-139 (December 1962).

l 19. The Effect of Nuclear Weapons. S. Glasstone. USAEC (June 1957).

l 20. Summary and Analysis of Accidents on Railroads in the U.S. - 1972. Accident Bulletin No.

141, U.S. Department of Transportation, Federal Railway Administration, Office of Safety (1973).

21. Telecon, Mr. R. W. Sharp, PGE, to Capt. D. E. Hughes, Columbia River Pilots Association (December 1973).

22. Telecon, Mr. J. L. Frewing, PGE, to Mr. Bob Olds, Crown Zellerbach, Wauna, Oregon (January 1974).

23. Telecon, Mr. J. L. Fewing, PGE, to Mr. Bob Boggess, Burlington Northern Railway (January 1974).

~

24. Letter, LT, Col. Edwin Opsted, Umatilla Army Ordnance Depot, Umatilla, Oregon, to J. E.

Grund, PGE (May 1973). '

25. U.S. NRC Regulatory Guide 1.78 (June 1974).

26. Telecon with Burlington Northern Railroad.

27. Columbia County, Oregon and Cowlitz County, Washington Maint Employer Profile Sheets, Chamber of Commerce (June 1992).

28. Telecon, Mr. R. N. Sherman, PGE, to Mr. Mike Dalich, Union Chemicals Division, Union Oil Company of California, Pasco, Washington (December 1980).

O Revision 8 2.6-4

29. Telecon, Mr. R. N. Sherman, PGE, to Mr. Ken Huitt, Hooker Chemical, Washington O

V (December 1980). ,

30. Telecon, Mr. R. M. Sherman, PGE, to Mr. Jeff Asay, Union Pacific Railroad Company, Portland, Oregon (December 1980).

31. Telecon, Mr. R. N. Sherman, PGE, to Mr. Thomas Ishman, Stauffer Chemical Company, Portland, Oregon (December 1980).

' 32. Telecon, Mr. R. N. Sherman, PGE, to Mr. C. H. Melcher, ACF Industries Inc., St Louis, Missouri (December 1980).

33. Telecon, Mr. R. N. Sherman, PGE, to Mr. Jeff Asay, Union Pacific Railroad Company, ,

Portland, Oregon (December 1980). i

34. Telecon with Union Chemicals Division, Union Oil Compiny.

35. Troian Nuclear Plant Final Safety Analysis Renort, Amendment 19 (December 1992). l REFERENCES FOR SECTION 2.3

1. Local Climarnlogien1 Data. Annual Summary with Comnarative Data. Portland. Oregon. l 1221. Environmental Data Services.

2. Climatic Atlac of the United States. ESSA, U. S. Department of Commerce (June 1968). l

3. Relation Between Gusts and A;.Trage Winds for Housing Load Determination, DGAI 140. l P. O Huss. Daniel Guggenheim Airship Institute, Arkon, Ohio, p 23 (June 1946).  !

4. Storm Data. Environmental Data Service, NOAA, U. S. Department of Commerce (issued l l

monthly) (1961-1972).

5. Weather Record Book United States and Canada. David M. Ludlum. Princeton: l Weatherwise Inc., p 98 (1971).

6. Low-Level Inversion Freauency in the Contituous United States. C. R. Hosler. Monthly l Weather Review, 89, 9; p 319-332 (1961).

7. Glossary of Terms Freauentiv Used in Air Pollution. V. J. Schaefer (Chairman). Boston: l American Meteorological Society, p 34 (1%8).

O '

2.6-5 Revison 8 J

l 8. Mixine Heights. Wind Soeeds. and Potential for Urban Air Pollution Throuchout the Conticuous United States, Preliminary Document. G. C. Holzworth. Environmental Protection Agency, p 146 (May 1971).

g l 9. A Study of Air Pollution Potential for the Western United States. G. C. Holzworth. Journal of Applied Meteorology,1,3; p 366-382 (1962).

l 10. Climatography of the United States No.10-45. Decennial Census of the United States -

Summary of Hourly Observations - Kelso-Longview. Washington. U. S. Department of Commerce (1970).

l 11. The Meteorological Program at the Trojan Nuclear Plant Site - 9/71 - 2/72, NUS-924. NUS Corporation (November 1972).

l 12. The Meteorological Program at the Trojan Nuclear Plant Site - 9/71 - 8/72, NUS-1002.

NUS Corporation (July 1973). -

l 13. The Meteorological Program at the Trojan Nuclear Plant Site - 9/71 - 8/73, NUS-1169.

NUS Corporation (July 1974).

l 14. Environmental Effects of Cooling Tower Operation at the Trojan Site - January 1969, NUS-510. NUS Corporation, p 6-27 (January 1969).

l 15. Evaluation of Envimnmental Effects of Cooling Tower Ooeration at the Trojan Site - May 19.22, NUS-911. NUS Corporation (May 1972).

l 16. Trojan Nuclear Plant Final Safety' Analysis Report, Amendment 19 (December 1992).

. REFERENCES FOR SECTION 2.4 l 1. Lower Columbia River - Standard Project Flood and Probable Maximum Flood. North Pacific Division Corps of Engineers (September 1969).

2. Department of the Army, Portland District, Corps of Engine; is.

3. Washington State Department of Water Resources, Olympia, Washington.

4. State Engineer, Water Rights Division, Salem, Oregon.

l 5. Hydrometeorological Tentative Estimates of Maximum Possible Flood-Producing Meteorological Conditions in the Columbia River Basin, Report No.18. U. S. Weather Bureau (January 1945).

Revision 8 2.6-6

s l 6.

O 3

Program Description and Ooerating Instnictions Streamflow Synthesis and Reservoir Regulations U. S. Army Engineer Division, North Pacific, Portland, Oregon (January 1964).

l

7. Applicatinn of Strenmflow Synthesis and Recervoir Regulation "SSARR" - Program to the l

' Lower Mekong River.' D. M. Rockwood. 'U. S. Army Engineer Division, North Pacific Division.

8. Chief Joseph Dam. Supplement 1 to Design Memorandum (Unnumbered). Spillway Design l Flood. U. S. Army Corps of Engineers, Seattle District (November 1%7).

9. livdrology Digital Model of Willamette Basin Tributaries for Onerational River Forecacting. l D. W. Kuehl and V. P., Schermerhorn (1%7).

10. Flood Control Ooerating Plan for Columhia River Trearv-Storage. International Task Force l on Flood Control Operating Plant Draft (August 12,1%8).

11. Computatinn of Freeboard Allowancec for Waves in Reservoirs, Technical letter l No.110-1-8. U. S. Army Corps of Engineers (August 1966).

12. Artificial Flood Considerations for Columbia PSrer Dams. U. S. Army Corps of Engineers [

.g (August 1%3).

V

13. Artificial Flood Considerations for Columbia River Dams. U. S. emy Cwr. of Engineers l )

(August 1%3).

14. Charles Goodwin Jr. letter (PGE) to' U. S. Nuclear Regulatory Cominission dated September 15,1980.

15. Analysis of Flood Irvel at Troian Nuclear Plant Associated with Hvoothetical Failure of l Spirit I *e Blockage. Simon, Li & Associates. Fort Collins, Colorado (August 15, 1983).

. 16. Oregon State University Marine Science Center, Newport, Oregon, private conununication.

17. Thermograph Charts for Trojan Thermograph, Portland, General Electric Company.

18. Onerating Rmtrictions at Bonneville Dam. Portland District Corps of Engineers l (July 26,1971).

19. A Study of the Plume from the Discharge Structure of the 1.130 MW(e) Troian Nuclear l Plant Design Reoort for Outfall Structure. S. R. Christensen and J. M. Mihelich. Portland General Electric Company, Portland, Oregon (1970).

2.6-7 Revison 8

l 20. Research Related to the Development of a Power Reactor Site on the bower Columbia River.

Battelle Memorial Institute, Pacific Northwest Laboratory, Richland, Washington (March 1,1969).

21. J. C. Sonnichsen, Battelle Northwest to S. R. Christensen, Portland General Electric Company, private communication (February 17, 1970).

l 22. Troian Nuclear Plant Final Safety Analysis Report, Amendment 19, Section 2.4 (December 1992).

l 23. Troian Nuclear Plant Final Safety Analysis Reoort, Amendment 19 (December 1992).

REFERENCES FOR SECTION 2.5 l 1. Ausnist 1972 Geophysical Survey Report - Troian Nuclear Power Plant Site. Trojan Geophysical Advisory Board.

l 2. Oregon Earthquakes - 1841 through 1958. J. W. Berg, Jr. and C. D. Baker. Bulletin of the Seismological Society of America,53,1 (1963).

l 3. A Summary of Washington Earthquakes. H. A. Coombs. Bulletin of the Seismological l Society of America,93,1 (1953).

\

l 4. Earthquake History of the United States, Revised ed. through 1963. R. A. Eppley. U. S.

Coast and Geodetic Survey (1965).

l 5. Washington State Earthquakes - 1940 through 1965. Norman Rasmussen. Balletin of the Seismological Society of America 57,3 (1967).

l l 6. Notices of Recent American Earthquakes. C. G. Rockwood. American Journal of Science l and Arts (1878).

l l l 7. Abstracts of Earthquake Renorts.1967 - September 1968, quarterly publication of the U.S. '

Department of Commerce, U.S. Coast and Geodetic Survey.

l 8. Description Catalog of Earthquakes of the Pacific Coast of the United States 1769 to 1928. S.

l D. Tomley and M. W. Allen. Bulletin of the Seismological Society of America,29, I (1939).

l 9. ESSA Earthouake Hypocenter Cards - Jan.1968 - Sept.1968. U.S. Department of Commerce.

Revision 8 2.6-8 a

10. Faults Peripheral to tne Troian Nuclear Plant. Oregon. H. A. Coombs (May 1972). l

11. Seismic Margin mrthouake For The Troian Site. May 1993. Project 1348, Geomatrix l-Consultants.

12. Geology and Mineral Resources of the Kelso-Cathimmet Area. Washington. l V. E. Livingston, Jr. Washington Division cf Mines and Geology Bulletin No. 54 (1966).

13. 1964 Geological Interoretation of Reconnaissance Gravity and Aeromagnetic Surveys in l Ronhwestern Oregon, USGS Bulletin 1181-N. R. W. Bromery and P. D. Snavely, Jr.

14. Geology of Portland. Oregon and Adiacent Areas, USGS Bulletin 1119. D. E. Trimble l

(1%3).

15. Volcanic Hm7mrds at Mt Rainier. Washington, USGS Bulletin 1238. D. R. Crandell and l

D. R. Mullineaux (1%7). -

16. Geologic Setting of Merrill Imke and Evaluation of Volcanic Hn7mrds in the Call River

[

Vallev Near Mt. St. Helene. Washington, USGS open-file report. J. H. Hyde (1970).

17. Thermal Features of Mt. Rainier. Washington as Revealed in Infrared Surveys, USGS l Professional Paper 525-D. R. M. Moxham, D. R. Crandell, and W. E. Marlatt.

p D93-D100 (1%5).

O. 18. 1 A New Steam Vent on Mt. Rainier. Washington. Z. F. Danes. Journal of Geophysical l l Research, 70, 8 (April 15,1%5).

19. Volcanic Havards in the Cascade Range. D. R. Crandell and H. H. Waldron. Geologic l Hazards and Public Problems, Conference Proceedings (May 27-28,1%9). (Sponsored by '

the Office of Emergency Preparedness, Region Seven, Federal Regional Center, Santa Rosa,

, California).

20. Volcanoes and Their Activity. A. Rittman, translated by E. A. Vincent. John Wiley & l Sons, New brk & London, Pitman Press, Bath (1962).

21. Mt. St. He!ans. A Recent Cascade Volcano. J. Verhoogen. University of California l PublMation, M, 9, p 263-302 (1937).

24 Volcanic Eruptiens The Pioneers' Attitude on the Pacific Coast from 1800 to IL75, l unpublished thesis. M. M. Folsom,1970. Department Pitman Press, Bath (1962).

25. Trojan Nuclear Site - Results oflaboratory Rock Testing. Bechtel Engineering Corporation l (undated).

2.6-9 Revison 8

l 26. Geophysical Survey of Troian Nuclear Power Plant Site - Longview Washincton.

P. C. Exploration (undated).

27. C. Goodwin letter (PGE) to T. M. Novak (NRC) dated September 15, 1980.

28. D. J. Broehl letter (PGE) to T. M. Novak (NRC) dated October 16,1980.

29. D. J. Broehl letter (PGE) to R. A. Clark (NRC) dated November 21,1980.

l 30. Peak Acceleration. Vebcity. and Displacement from Strong-Motion Records. D. M. Boore, W. B. Joyner, A. A. Oliver III, and R. A. Page. Bull. Seismological Soc. Amer. 70(1), p 305-321 (1980).

l 31. Intensity of Ground Shaking near the Causative Fault. G. W. Housner. Proc. of the Third .

World Conf. Earthquake Eng., New Zealand (1965).

l 32. Preliminnry Annivsis of the Peaks of Strong Fnrthqunke Ground Motion - Dependence of Peaks on Earthquake Magnitude. Epicentral Distance and Recording Site Conditions.

M. D. Trifunac. Bull. Seismological Soc. Amer., 66,1, p 189-220 (1976).

j 33. Seismic Regionalization Studies Bonneville-Power Service Area. Washington. Oregon. Idaho and Western. R. W. Couch and R. J. Deacon. Shannon & Wilson, Inc., report to Agbabian 1 Associates (1972).

l 34. Geologic Map of the St. Helens Ouadrangle. W. D. Wilkinson, W. D. Lowry, and E. M. Baldwin. State of Oregon,' Department of Geology and Mining Industry, Bulletin No.

31 (1946).

l 35. United States Earthquakes - 1928 - 1968. Annual publication of the U.S. Department of Commerce, U.S. Coast and Geodetic Survey.

l 36. A Catalogue of Earthquakes on the Pacific Coast - 1769 to 1897. E. S. Holden.

l 37. Troian Nuclear Plant Final Safety Analysis Report, Amendment 19 (December 1992).

Revision 8 2.6-10 u

i.

4-' 3.0 FACILITY DESIGN Chapter 3 discusses the design of facilities and systems that support the storage of spent fuel in j the Spent Fuel Pool at the Trojan Facility.

l 9

9 e

e l

I O 3.0-1 Revision 8

3.3 AUXILIARY SYSTEMS This section discusses the auxiliary systems that are used to support the storage of spent fuel at Trojan. This section includes discussions on the Fuel Handling System Modular SFP Cooling and Cleanup System, Service Water System, Compressed Air System, Makeup Water Treatment System, equipment and floor drain systems, Plant discharge and dilution stmeture, Primary Sampling System, Fire Protection System and Program, Control Room habitability, and seismic instrumentation. These systems do not perform any safety functions with the reactor defueled.

I 3.3.1 FUEL HANDLING SYSTEM 1

The Fuel Handling System consists of equipment and structures utilized for handling spent fuel assemblies during fuel transfer operations. This discussion is limited to fuel handling equipment

]

used for transfer operations within the SFP. A description of fuel handling equipment used for j transfer operations between the SFP and Transfer and Storage Casks in the Fuel Building is l l provided in the ISFSI SAR, LCA 237, r.nd LCA 246. Transfer of fuel'to the Containment l Building is prohibited under Trojan's current license. l 3.3.1.1 Design Bases The Fuel Handling System is designed to minimize the possibility of mishandling or maloperation that could cause fuel damage and potential fission product releases. The following design bases apply to the Fuel Handling System:

(1) Fuel handling devices have provisions to avoid dropping orjamming of fuel e>semblies during transfer operations.

O' 3.3-1 Revision 8 1

(2) The fuel handling equipment has been designed for the loading that would occur during a Safe Shutdown Earthquake (SSE). The fuel handling equipment will not fail so as to cause damage to any fuel elements should a SSE occur during fuel transfer operations.

(3) The hoist used to lift the spent fuel assemblies has a limited maximum lift height which is determined by the length of the long-handled tool, so that the minimum required depth of water shielding is maintained.

Environmental conditions of the fuel handling equipment, such as exposure to borated water and high humidity,' are considered in the design and selection of the material.

3.3.1.2 System DescriptiRD 3.3.1.2.1 General Description Fuel assemblies are moved in the SFP using the SFP bridge hoist. When lifting spent fuel assemblies, the hoist uses a long-handled tool to assure that sufficient radiation shielding is maintained. Fuel assembly inserts, such as thimHe plugs, burnable poisons rods, rod control clusters, and source rods, may also be transferred between positions within the SFP.

3.3.1.2.2 Component Description

( 3.3.1.2.2.1 Fuel Buildine bridee crane. The Fuel Building bridge crane is an indoor electric overhead travelling bridge crane complete with a single trolley and all the necessary motors, control,

< brakes, and control station. The main hoist of the crane is rated at 125 tons and the auxiliary hoist at l

l 25 tons. The crane and accessories have been designed and constructed for indoor service and were l

l desigt = u to handle new and spent fuel containers between the railroad cars and loading and l i

unloading pits. Movement of i

1 3.3-2 l

l

In the event of a design basis seismic event, the systems that were designed to Seismic Category II

/ requirements may not be operable. In this case, sufficient time would be available to align a source b of makeup water to maintain water level; 3.3.3 Deleted l 3.3.4 SERVICE WATER SYSTEM The SWS is designed to provide water from the Columbia River to supply water to various systems l and equipment. With the reactor permanently defueled, the primary function of the SWS is reduced l te providing makeup to the SFP. , l 3.3.4.1 Design Bases 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 permanently O) defuel L"

e, d emsystdesign requirements are reduced substantially from the original design bases of the system.

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

l

/m bl 3.3-9 Revision 7

3.3.4.2 System Description The components included in the Service Water System are:

(1) Three SWS pumps (P-108A, B, and C). Note that P-108A has been rebuilt. It may remain removed from the system indefinitely l

l (2) Interconnecting piping, valves, and instrumentation  ;

I i

l (3) Two cooling water pumps (P-167A and B)

I Water is supplied through the trash rack and traveling water screens at the intake structure. The l water entering the system can be chlorinated for microbiological control.

One traveling water screen is installed in the flow path to each of the two independent flow paths to the service water pumps at the intake structure. The screens are automatically cleaned by the screen wash system which consists of two vertical pumps taking suction on the downstream side of the screens and discharging into high velocity. spray nozzles 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.

i 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. Power to P-108A and P-108C is supplied from 4160-V bus Al and P-108B is l supplied from 4160-V bus A2. Lubricating water is supplied to the three service water pumps.

River water can also be supplied to the SWS using Coohng Water Pumps P-167A and P-167B.

P-167A can supply makeup flow to the SFP.

I l Bearing lube water can be supplied directly from the service water pumps.

Revision 8 3.3-10 O

- 1 4

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

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.

3.3.5 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 syste'm 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 headeris connected to the air-dryer unit discharge. Each air header supplies branch lines which supply l instrument air and service air to the required loads throughout the Plant. The instrument and service air system provides air to the inflatable seals for the SFP gates.

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.

3.3.7 Deleted

{

A V 3.3-11 Revision 8

3.3.8 EOUIPMENT AND FLOOR DRAIN SYSTEMS . ,

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

(1) Liquid Radioactive Waste Treatment System (LRWS) drains (2) Oily Waste System (3) Acid Waste System (4) Sanitary Waste System ._

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

The LRWS is designed to collect liquid waste from areas contaming equipment that handles radioactive fluids. This system is designed to control the spread and release of radioactive particulates by directing potentially radioactive fluids to the Radioactive Waste Treatment System.

This system is described in Section 5.3.

The Oily Waste System collects the waste from floor 'and equipment drains in places where the waste is not potentially radioactive. The waste is conveyed to a settling tank where the oil is separated prior to releasing the water to the Plant discharge and dilution structure.

The Acid Waste System is designed to drain fixtures and equipment in which chemicals are expected to be present in nonradioactive effluent. Selection of piping materials was based on providing a surface resistant to corrosion. The waste is drained into the neutralizing tank, T-126, where it is neutralized. The neutralized waste is then transferred to the Plant discharge and dilution structure or solid settling basin using the neutralizing drain tank pumps.

Revision 8 3.3-12

The Sanitary System collects waste from floor drains in toilet rooms, shower rooms not requiring radioactive waste connections, and the plumbing fixtures.

3.3.9 PLANT DISCHARGE AND DILUTION STRUCTURE The Plant discharge and dilution structure receives the Plant liquid radioactive and chemical wastes, provides dilution water, and discharges the diluted waste to the Columbia River. The residual chlorine concentration of the discharge is controlled by the addition of sodium bisulfite at the Plant

{

i effluent. Diluted chemical waste discharge concentrations are diluted to meet the requirements of i the State of Oregon Department of Environmental Quality. The structure does not perform any safety functions.  ;

3.3.10 Deleted 3.3.11 FIRE PROTECTION SYSTEM AND PROGRAM

~

O Fire Protection is provided for the Plant as described in the Topical Report, " Trojan Nuclear Plant Fire Protection Program", PGE-1012.

?!

- 3.3.12 CONTROL ROOM HABITABILITY To support habitability, the original Control Room design included radiation shielding, air filtering, 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 Revision 8

3.3.13 SEISMIC INfiTRUMENTATION Seismic instrumentation for the facility consists of a multielement seismoscope and peak acceleration recorders. The multielement seismoscope is mounted on an essentially rigid foundation, which will provide no significant amplification of ground motion. Peak acceleration j

f recorders are also installed at appropriate locations in the facility. The seismic instrumentation I 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.

O 5

Revision 8 3.3-14

3.6 REFERENCES

REFERENCES FOR SECTION 3.1

1. Troian Nuclear Plant Final Safety Analysis Reoort, Amendment 19 (December 1992). l

2. Seismic Analyses of Structures and Eauipment for Nuclear Power Plants, BC-TOP-4. Bechtel l Corporation (April 30,1971).

3. Seismic Analyses of Structures and Eauipment for Nuclear Power Plants, BC-TOP-4A, l Revision 3. Bechtel Power Corporation (November,1974).

l

4. PGE-1.020. Renort on Design Modifications for the Troian Control Building, Revision 4, l Portland General Electric Company (February 12,1980).

l

5. Ground Motions. Trojan Nuclear Plant. Rainier. Oregon. I. M. Idriss (November 1970). l

6. Seismic Design Criteria for Nuclear Power Plants. Newmark and Hall. Presented at the Fourth l World Seismic Conference held at Santiago, Chile (1969).

7. Building Foundation Interaction Effect. R. A. Parmelee. Eng. Mech. Div. Spec. Conf. ASCE l (October 12,1966).

8. A' Renlacement for SRSS Method in Seismic Analysis. E. L. Wilson, A'. Der Keureghian, and l 1 E. P. Bayo. Earthquake Engineering and Structural Dynamics, Volume 9 (1981).

l

9. PGE-1052. Qualitv-Related List Classification Criteria for the Trojan Nuclear Plant, l Amendment 2 (August 1993).

l REFERENCES FOR SECTION 3.2

1. Troian ' Nuclear Plant. Final Safety Analysis Reoort, Amendment 19 (December 1992).

2. Assessment of Containment Structure Post-Tensioning System Surveillance Program. l J. K. McCall, G. Ferrell, and J. V. Rote. Bechtel Power Corporation (May 1973).

3. PGE-1020. Renort on Design Modifications for the Trojan Control Building, Revision 4 l (February 12,1980).

3.6-1 Revision 8

4. PGE letter to A. Schwencer dated June 29,1979, Response to NRC Staff Question 11(a) dated May 18,1979.

l 5. Evaluation of Tensile Bond and Shear Bond of Masonry by Means of Centrifugal Force.

M. .Hatzinikolas, J. Longworth, and J. Warwaruk. Alberta Masonry Institute, Edmonton, Alberta.

l 6. Cvelic Loading Tests of Masonry Single Piers - Volume 3. CB/EERC-79-112. P.A. Hidalgo, R. L. Mayes, H. D. McNiven, and R. W. Clough. College of Engineering, University of Califomia, Berkeley, California (May 1979).

l 7. Cyclic Loading Tests of Masonry Single Piers - Volume 2 - Height to Width Ratio of 1, Report Number UCB/EERC-78-28. S. J. Chen, P. A. Hidalgo, R. L. Mayes, R.W. Cough, and H. D. McNiven. College of Engineering, University of California, Berkeley, California (December 1978).

l 8. Bechtel Ouality Assurance Program for Nuclear Power Plants. BQ-TOP-1, Revision 2A.

Bechtel Engineering Corporation (July 1977).

l 9. Licensee's Testimony on Matters Other than Structural Adeauacy of the Modified Complex.

Prepared by Broehl, et.al. (March 17,1980).

l 10. PGE-1037. Trojan Nuclear Plant Soent Fuel Storage Rack Replacement Report. (July 1983).

Historical only.

l 11. Trojan Operating License. No. NPF-1, Amendment No.196 (May 19,1997).

12. Letter, B. D. Withers (PGE) to L. Franli(ODOE), dated April 17,1984.

REFERENCES FOR SECTION 3.3

1. Trojan Nuclear Plant Final Safety Analysis Report, Amendment 19 (December 1992).

l 2. Decay Heat Power in Nuclear Reactors. American Nuclear Society Standard ANSI /ANS 5.1-1979, approved by ANS-5.

l l l 3. PGE-1069. Troian Indeoendent Soent Fuel Storage Installation Safety Analysis Reoort.

l l l 4. License Change Acolication (LCA) 237. Soent Fuel Cask Loading in the Fuel Building.

l Transmitted by PGE letter, VPN-005-99, dated January 7,1999.

Revision 8 3.6-2 O

I i l

l

1 I

4

5. LCA 246. Spent Fuel Cask Loading in the Fuel Buildino - Contingency Fuel Unloading to the l Fuel Pool. Transmitted by PGE letter, VPN-006-99, dated January 27,1999.

O l

REFERENCES FOR SECTION 3.4 l

1

1. Troian Nuclear Plant Final Safety Annivsis Reoort, Amendment 19 (December 1992).

I REFERENCES FOR SECTION 3.5

1. PGE-1028. Troian Nuclear Plant Regulatorv Guide Poliev Manual, Amendment 13 l.

(March 1992).

O .

)

)

1 I

9 3.6-3 Revision 8 1

TABI F 3.5-1 Sheet 1 of 35 j I

9 1 LIST OF PERTINENT REGULATORY GUIDES FOR DEFUELED CONDITION Revision and Compliance Status l Regulatory Guide j

1.8 - Personnel Selection and Training (5/77), Rev.1-R Comnliance Statue I 1

- Comply with exception. -

i (a). Technical Specifications require that each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 except for the Radiation Protection Manager, who shall meet or exceed the qualifications of Regulatory Guide 1.8, April 1987, and Independent Safety Reviewers, who shall have five years of professional level experience and either a Bachelor's Degree in Engineering or the Physical Sciences or equivalent in accordance with ANSI /ANS-3.1-1981.

-(b) The qualification requirements of ANS 3.1-1981 are used as a goal for the Technician ,

with one exception. Specifically the Technician requirements given in ANSI N18.1-- l 19_71 (2-year working experience and an optional 1 year of related technical training) {

will be used as a minimum requirement in lieu of the 3 years experience required for ANS 3.1-1981.

. .(c) Paragraph 4.2.2 of ANSI N18.1-1971 requires there be an operations manager who holds a Senior Reactor Operator (SRO) license. Paragraphs 4.3.1 and 4.5.1 discuss supervisors and operators required to have NRC licenses. Section 5.2 discusses training of candidates for NRC examinations. Paragraph 5.5.1 recommends retraining include startup and shutdown procedures and emergency shutdown systems. T lieu of the above, no personnel are required to hold NRC licenses, and Certified Fuel Handler (CFH) training will only be given on subjects applicable to the permanently defueled mode. In addition, CFH training need only maintain proficiency in allowed activities.

Revision 8 1

TABLE 3.5-1 Sheet 2 of 35 Regulatory Guide

- j 1.13 - Fuel Storage Facility Design Basis (3/71), Rev. 0 l Compliance Status Comply with exception.

A seismic Category I makeup system is not provided. The reduced decay heat load after 3 %

years since the reactor was defueled allows over 10 days with no makeup or cooling before manual actions are required to prevent fuel damage. Therefore, adequate time is available to operate or restore one of the diverse non-seismic makeup sources.

Rerulatory Guide -

1.21 - Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants (6/74), Rev.1 j Compliance Starne Comply with exception. l The use of a meteorological tower for reai time measurement of meteorological conditions may not be used. Worst case meteorological values, based on previous evaluations, will be used to conservatively estimate releases when the meteorological tower is not used.

Regulatory Guide 1.24 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Gas Storage Tank Failure (3/72), Rev. O Compliance Status Comply.

Revision 8

TABLE 3.5-1 Sheet 3 of 35 A

Regulatory Guide 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (3/72), Rev. O Compliance Status Comply with exception.

The assumptions used in this Regulatory Guide for fuel rod gap fractions and iodine pool decontamination factors are based on certain fuel performance limits. One of the assumed limits of assembly average burnup < 25,000 mwd /MTU is lower than expected for Trojan

- fuel. Specifically, Trojan fuel average burnups can be as high.as 44,000 mwd /MTU.

However, an analysis showed that the gap fractions in the Regulatory Guide still conservatively bound realistic case calculations up to 44,000 mwd /MTU. Therefore, the assumptions of this Regulatory Guide are still valid for Trojan fuel.

. Revulatory Guidet 1.26 - Quality Group Classifications and Standards-for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plants (2/76), Rev. 3 Compliance Comply with exception.

Safety-related structures, systems, and components were originally classified into quality groups prior to the issuance of Revision 0 to Regulatory Guide 1.26. Currently, the facility classification process is described in PGE-1052, " Quality-Related List Classification Criteria for the Trojan Nuclear Plant", for the permanently defueled condition.

t

l TABLE 3.5-1 Sheet 4 of 35 1 __._

Regulatorv Guide 1.28 - Quality Assurance Program Requirements (Design and Construction) (8/85), Rev. 3 Compliance Comply with exception.

The Trojan Nuclear Plant was initially designed and constructed in accordance with the Quality Assurance Program included in Chapter 17 of the FSAR. This program contained those measures necessary to assure adequate quality in the completed facility in accordance with Appendix B to 10 CFR 50. Subsequent and future modifications to the Trojan Nuclear Plant are carried out under the provisions of Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation), (Revision 2)" and the guidelines of PGE's Nuclear Quality Assurance Program (PGE-8010).

Reculatorv Guide i

1.29 - Seismic Design Classification (9/78), Rev. 3 Compliancs Comply with exception.

(a) Regulatory Positions C.1.d and C.1.g require that the cooling water and systems or portions of systems that are required for cooling the spent fuel storage pool be designed as Seismic Category I. However, the Modular Spent Fuel Pool Cooling and Cleanup System has been designed to Seismic Category II requirements. In the current defueled condition, the reduced decay heat available to heat-up the spent fuel pool allows adequate time to supply water for makeup due to boiling from numerous non-seismic sources. This diversity ensures adequate protection of the spent fuel.

(b) Footnote 2 to Regulatory Position C.1.p states that specific guidance on seismic requirements for radioactive waste management systems is under development.

Seismic requirements for radioactive waste management systems are addressed in Regulatory Guide 1.143 (Rev.1, October 1979). Regulatory Guide 1.143 requires the foundations and walls of structures housing the liquid and solid radwaste systems to be Seismic Category I, and permits the equipment used to collect, process, and store liquid and solid radioactive wastes to be Seismic Category II. Regulatory Guide 1.143 Revision 8

TABLE 3.5-1 Sheet 5 of 35 O

also requires portions of gaseous radwaste. treatment systems that are intended to store or delay the release of gaseous radioactive waste, including portions of structures housing the system, to be Seismic Category I. Whereas Trojan's seismic classification guidelines for radioactive waste management are not as specific as those given in Regulatory Guide 1.143, the seismic classification of Trojan's solid, liquid, and l gaseous radwaste systems, and the structures housing the systems, are in full compliance with the above seismic requirements of Regulatory Guide 1.143 for such

systems-.

(c) Regulatory Position C.I.p of Revision 3 to Regulatory Guide 1.29 also requires that any other systems (other than radwaste systems) not et = red specifically in the Regulatory Guide that contain or may contain radioactivv material, and whose postuisted failure would result in conservatively calculated potential offsite doses greater than 0.5 rem (whole body or its equivalent to any part of the body), be designated Seismic Category I. The Trojan classification method does not contain this additional requirement for the classification of structures, systems, and components outside of the Control, Auxiliary, and Fuel Building Complex. These items were classified in accordance with Revision 0 of Regulatory Guide 1.29.

(d) Regulatory Position C.2 of this Regulatory Guide requires that portions of non-safety-related systems whose failure could reduce the functioning of a Seismic Category I plant feature to an unacceptable level be designed and constructed so that the SSE would not cause such failure. The original Plant design included system interaction considerations as well as failure modes and effects analyses primarily in determining system and equipment locations. Portions of non-safety-related Seismic Category II systems (e.g., pipe supports) were not originally designed to Seismic Category I requirements.

The system interaction requirements in this Regulatory Guide are now implemented in Seismic II/I provisions. A system interaction review identifying potential II/I items has been completed for safety-related systems and equipment in the Control, Auxiliary, and Fuel t

Buildings. Design and analysis of Seismic Category II/I items are for SSE seismic loadings (operating basis earthquake loads are not required to be analyzed). The design and construction of Seismic Category II/I items will be considered quality-related but not safety-related.

The original Plant design meets Regulatory Position C.2 that is given in Revision 0 to this Regulatory Guide. The position to design and construct Seismic Category II/I items for SSE loads complies with Revision 3 to this Regulatory Guide. .

Revision 8

t TABLE 3.5-1 Sheet 6 of 35 Renulatory Guide 1.30 - Quality. Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electrical Equipment (8/72), Rev. O Comnliance

. Comply.

Renulatorv Guide 1.33 - Quality Assurance Program Requirements (2/78), Rev. 2 Comnliance -

Comply with exception.

(a) Section 4.3.1 of ANSI N18.7-1976 requires personnel assigned responsibility for 7 independent reviews to collectively have the experience and competence required to review problems in the following areas: nuclear power plant operations; nuclear engineering; chemistry and radiochemistry; metallurgy; nondestructive testing; instrumentation and control; radiological safety; mechanical and electrical engineering; administrative controls and quality assurance practices; and other appropriate fields associated with the unique characteristics of the nuclear power plant involved.

In lieu of the above, as specified in Technical Specification 5.5.2, personnel assigned to the Independent Review and Audit Committee shall collectively have experience and knowledge in the following functional areas: fuel handling and storage; chemistry and radiochemistry; engineering; radiation protection; and quality assurance.

(b) Section 4.3.4(2) of ANSI N18.7-1976 requires the independent review body to review

" Proposed changes in procedures, proposed changes in the facility, or proposed tests or experiments, any of which involves a change in the technical specifications..." PGE meets the intent of this requirement via the Independent Review and Audit Committee's review of the associated technical specification change.

l (c) Section 5.1 of ANSI N18.7-1975 requires a summary document of those sources providing administrative controls and quality assurance during the operational phase to index such source documents to the criteria of ANSI N18.7-1976. PGE has not Revision 8 l t

]

TABLE 3.5-1 Sheet 7 of 35 compiled such a summary dec:rnent that indexes procedures and instructions to this sta'idard since the manpower required to perform such a task is notjustified.

(d)

Paragraph 5.2.1.1 of ANSI N18.7-1976 discusses the RO authority and responsibility.

Paragraph 5.2.1.3 discusses the SRO responsibilities. No personnel are required to hold RO or SRO licenses per the Technical Specifications.

(e)-

Section 5.2.2 of ANSI N18.7-1976 requires temporary changes which do not change the intent of an approved procedure to be approved by the supervisor in charge of the shift who holds a SRO license on the unit affected. Temporary procedure changes which do not change the intent shall be approved by a member of Plant Management and by a Certified Fuel Handler. This approved temporary change is documented, reviewed, and approved within 14 days by the responsible manager in accordance with approved administrative procedures. -

(f) Not all quality,-related procedures are reviewed biennially.

The following non-routine procedures whose usage is dictated by a facility event shall be reviewed at least every 2 years to determine if changes are necessary or desirable:

Off-NormalInstructions, Alarm Response Procedures, Emergency Procedures for the Radiological Emergency Plan and Nuclear Security Safeguards Contingency Procedures.

Other Plant procedures shall be reviewed and revised appropriately when affected by a corrective action, a change in the license or Technical Specifications, an update to the Safety Analysis Report, a Topical Report, A Plant Setpoint Change, a Vendor

' Equipment Technical Manual change, or a similar action required by Section 5.2 of ANSI N18.7-1976, i.e., any modification to a system, an unusual incident (accident),

unexpected transient, significant operator error or equipment malfunction. These procedures will be reviewed and audited as specified in procedures, but not necessarily every two years. These measures will ensure that procedures are being revised when appropriate and that the procedure review and revision program is effectively implemented.

(g)

Section 5.3 of ANSI N18.7-1976 requires procedures for starting up the reactor, steady-state power operation and load changing, shutting down and tripping the reactor, and changing modes. Procedures are not required for these actions in a defueled mode.

Revision 8

TABLE 3.5-1 Sheet 8 of 35 (h)- ANSI N18.7-1976 states that certain other ANSI standards will be utilized for compliance with ANSI N18.7-1976. PGE complies with these standards and their associated Regulatory Guides only as described in each Regulatory Guide position in the DSAR.

(i) Section 5.2.6 of ANSI N18.7-1976 requires procedures "... for control of equipment, as necessary, to maintain personnel and reactor safety and avoid unauthorized operation of equipment. The procedures shall require independent verification, where appropriate, to ensure that necessary measures, such as tagging equipment, have been implemented correctly."

PGE procedures do not require independent verification for clearances far equipment removed from service or system lineups except for designated locked valves. Trojan will maintain personnel safety by performing clearance operations in accordance with Occupational Safety and Health Administration (OSHA) regulations.

PGE procedures do not require independent verification for syste;n alignments performed in conjunction with operating procedures. Problems resulting from component mispositioning can be identified and corrected within sufficient time to prevent fuel damage. Due to plant closure, reactor safety is no longer a concern.

Regulatorv Guide 1.37 - Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants (3/73), Rev. 0 -

Compliance Comply.

p

~

TABLE 3.5-1 Sheet 9 of 35 V b -

Regulatory Guide 1,38 - Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling ofItems for Water-Cooled Nuclear Power Plants (5/77), Rev. 2 Compliance Rev. 2 - Comply with exception.

(a) \

Quality Assurance personnel auditing inspection activities will be qualified per ANSI N45.2.23 in lieu of ANSI N45.2.6 l (b)

The requirements of ANSI N45.2.2-1972 for classification levels will be applied to quality-related materials, parts, and components only when specific protection measures are required. In lieu of the detailed requirements of ANSI N45.2.2-1972, quality-related items will receive a thorough engineering evaluation to assure that adequate protective measures are specified for packaging, shipping, receiving, storage, and handling of items for nuclear power plants. These protective measures will be consistent with standard / commercial engineering practices and manufacturers'

.( recommendations.

(c) Marking may be applied to bare austenitic stainless steel and nickel alloy metal surfaces provided that it has been established that the marking is not deleterious to the item rather than as stated in Paragraph A ' 3.9 of ANSI N45.2.2. Proper chemical controls

-will be employed to ensure there is no adverse metallurgical impact on the steel or nickel alloy. This position has been adopted in the document ASME NQA-2, Addenda 2a.

1 Regulatorv Guide \

1.39 - Housekeeping Requirements for Water-Cooled Nuclear Power Plants (9/77), Rev. 2 '

Compliance Comply with exception.

1 Regulatory Guide 1.39 endorses the application of the recommendations of ANSI Standard N45.2.3, " Housekeeping During the Construction Phase of Nuclear Power Plants." The cleanliness and housekeeping standards established in ANSI Standard N45.2.3 are designed to control work activities, conditions, and environments that can affect the quality Revision 8

F TABLE 3.5-1 Sheet 10 of 35 v)

• of important parts of a nuclear power plant. These parts include structures, systems, or components whose satisfactory performance is required for the plant to operate reliably, to prevent accidents that cause risk to the health and safety of the public, or to mitigate the consequences of such accidents if they were to occur. Specifically, the last paragraph of Section 2.1, " Planning " requires a written record of the entry and exit of material be established and maintained for Zone III cleanliness areas. In addition, the second paragraph of Section 3.2, " Control of Facility," states that appropriate control measures shall be provided through the utilization of such items as log books and tethered tools.

In that the plant no longer operates and has a relatively long time available to respond to, and mitigate the consequences of, any accidents postulated in this DSAR, strict application of the requirements of ANSI N45.2.3 is no longer necessary. Specifically, the use of a log for control of material above the spent fuel pool will no longer be required, but the option for its use will not be precluded. If a log for control of material is not used, a visual inspection of the tops of spent fuel assemblies will be performed.

Regulatorv Guide

(

1.58 . Qualification of Nuclear Power Plant Inspection, Examination and Testing Personnel (9/80), Rev.1 Withdrawn Compliance Comply with exception.

Position C.2 requires certification of non-destructive examination (NDE) personnel per SNT-TC-1 A-1975. The NDE personnel are certified per SNT-TC-1 A-1984 at the Trojan Facility.

Regulatory Guide 1.59 - Design Basis Floods for Nuclear Power Plants (8/77), Rev. 2 Compliance Comply with exception.

During the design phase of Trojan, pertinent regulations dealing with floods were 10 CFR 100.1 and General Design Criterion 2. The probable maximum flood (PMF) for Revision 8

TABLE 3.5-1 Sheet 11 of 35 O

v .

Trojan was determined prior to the issuance of this Regulatory Guide and has been reviewed and approved by the NRC. As stated in this Regulatory Guide, previously reviewed and approved detailed PMF analysis at specific sites will continue to be acceptable.

x Reentatnry Guide 1.60 - Design Response Spectra for Seismic Design of Nuclear Power Plants (12/73), Rev.1 Comnliance Comply with exception.

Design response spectra for the OBE and SSE were derived and used in the seismic analysis and design of the Trojan Nuclear Plant prior to issuance of this Regulatory Guide. The methods used in the original derivation are described in FSAR Section 3.7.1 and have been reviewed and approved by the NRC. DSAR Section 3.1 5 summarizes the seismic design parameters.

• Reenlatorv Guide 1.61 - Damping Values for Seismic Design of Nuclear Power Plants (10/73), Rev. 0 Compliance Comply with exception.

(a) Table 1 of this Regulatory Guide specifies a damping value from 1 to 3 for piping systems (dependent upon pipe diameter and type of earthquake). PGE used a critical i damping value of 0.5 percent in the dynamic analysis of vital piping systems and '

equipment. Seismic analysis of Trojan piping systems using a damping value of 0.5 percent is more conservative than the criteria specified by this Regulatory Guide.

(b) Table 1 of this Regulatory Guide specifies a damping value of 4 and 7 percent for an OBE and SSE, respectively, for reinforced concrete structures. PGE used critical damping values of 2 and 5 percent for the OBE and SSE, respectively. Analysis of Trojan's Category I structures employing damping values of 2 and 5 percent is more conservative than the criteria specified by this Regulatory Guide.

Revision 8

TABLE 3.5-1 Sheet 12 of 35 (c) Principle constituents of the Control, Auxiliary, and Fuel Building complex are reinforced concrete, reinforced grouted masonry, and a combination of the two. Table 1 of this Regulatory Guide does not specify damping values for such heterogeneous structures. However, the complex's structural materials can be conservatively assumce analogous to reinforced concrete for damping considerations. Damping values of 2 and i 5 percent were used for an OBE and SSE, respectively, and are adequate.

l (d) Table 1 of the Regulatory Guide specifies damping values of 2 to 3 percent for equipment,2 to 4 percent for welded structures, and 4 to 7 percent for bolted structures. For structures and equipment, and supports for other than vital piping systems, PGE used damping values higher than 0.5 percent where justified. A -

maximum of 2 percent for OBE load' combinations and 5 percent for SSE load combinations was used by PGE in accordance with the original design approach.

Alternatively, some damping ratios are based on calculated stress levels as provided in DSAR Table 3.1-5. The use of 2 percent for OBE and 5 perce' t for SSE, where justified, or damping ratios based on stress levels as provided in DSAR Table 3.1-5 meets the intent of the Regulatory Guide and is adequate.

p Regulatorv Guide -

d 1 i

1.64 - Quality Assurance Requirements for the Design of Nuclear Power Platits (4/76), Rev. 2' i Withdrawn l

Compliance '

Comply with exception.

Under certain conditions, the Manager, Engineering / Licensing may perform design verification provided:

(a) This manager is the only technically qualified individual or has special technical expertise which would allow him to perform a more thorough design verification.

(b) The need is individually documented and approved in advance by the General Manager, Trojan Plant.

(c) Quality Department audits cover frequency and effectiveness of use of the Manager, Engineering / Licensing as a design verifier to guard against abuse.

TABLE 3.5-1 Sheet 13 of 35 i

g This position is consistent with NUREG-0800.

Reenlarnrv Guide j 1

1.74 - Quality Assurance Terms and Definitions (2/74), Rev. 0 WitMrawn l l

Comnliance Comply with exception.

The terms and definitions of ANSI N45.2 daughter standards that are not in agreement with ANSI N45.2.10 will be as used in the daughter standard. In addition, the definitions of Permanent (Lifetime) Records and Nonpermanent Records have been modified for clarity in accordance with the definition in the QA Program Glossary. The definition of " repair" in ANSI N45.2.10 is replaced by " modify", because the term " repair" as used in various sections of the ASME Boiler etPressure Vessel Code has a different mea mg. The term " modify" is used to avoid confusion.

l Reguintory Guide O 1,76 - Design Basis Tornado for Nuclear Power Plants (4/74), Rev. O Comoliance , ,

Comply with exception.

. Design, construction, and licensing activities for the Trojan Nuclear Plant were in an advanced stage prior to issuance of this Regulatory Guide. The recommendations of this Regulatory Guide were therefore not used as a basis for characterizing a design basis tornado. For the Trojan Nuclear Plant, criteria to evaluate safety-related structures.for potential effects of tornados were developed pursuant to AEC General Design Criterion 2. All Plant structures containing systems needed to achieve and maintain safe shutdown were determined to be capable of resisting 200-mph tornado loads and a pressure differential of 1.5-psi bursting pressure applied at 1 psi /sec. Many of these structures were evaluated to be further capable of resisting 300-mph tornado loads and a pressure differential of 3-psi bursting pressure at 1 psi /sec. Draft Standard ANS 2.3 specifies a 200-mph tornado at a probability of 10 yr' for sites in Trojan's region. The 200-mph and 300-mph tornado maximum wind speed criteria are conservative with respect to the wind speed associated with the maximum tornado that could reasonably be hypothesized for the site. These criteria and methods have been reviewed and Revision 8 a

t

i TABLE 3.5-1 Sheet 14 of 35 ID V -

approved by the NRC.

Regulatory Guide 1.78 - Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release (6/74), Rev. O i

Compliance ,

I Not applicable.

I Analysis shows that Control Room habitability is not a conceri$ in the defueled configuration. l See the discussion in Section 3.3.12 of the DSAR. .

l Regulatory Guide 1.88 - Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records (10/76), Rev. 2 Withdrawn g)

( Compliance Comply with exception.

I (a) The QA records will be stored in a facility with a minimum 2-hour fire rating in l' accordance with ANSI N45.2.9-1979, or as NFPA Class I records in a Class B (350-2 hours) container in accordance with NFPA 232-1975. ANSI N45.2.9-1979 has 3 superseded ANSI N45.2.9-1974 and requires a minimum fire rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for i facilities in which QA records are stored. NFPA 232-1975 is endorsed in this Guide as an acceptable alternative to the fire protection provisions listed in Subdivision 5.6 of

{

ANSI N45.2.9-1974.

l (b) The retention requirements of ANSI N45.2.9-1979 will be followed in lieu of the requirements in ANSI N45.2.9-1974. The retention requirements of  ;

ANSI N45.2.9-1979 appropriately reference the requirements of the regulatory agencies having jurisdiction over these records. {

i (c) Trojan will use optical storage as allowed by NRC Generic Letter 88-18, " Plant l

Record Storage on Optical Disks."

^

G S ..

Revision 8 i

.s i

l

TABLE 3.5-1 Sheet 15 of 35 Reguintorv Guide .

1.91 - Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants (2/78), Rev.1 ,

Comnliance Comply with exception. )

An exception is taken to the value of 1 psi for the maximum permissible overpressure generated by the atmospheric shock from an explosion. The analysis given in DSAR Section 2.2.3.1 calculated the probability of occurrence of a shock wave with a 2.2-psi overpressure. Test structures similar to the safety-related structures at Trojan have demonstrated that overpressures of 5 psi can be endured without any significant damage to these structures. Furthermore, it is noted that these structures are designed to withstand pressure differentials of a least 3 psi due to tornado effects. For credible accidents, however, the maximum permissible overpressure value used in the analyses of Trojan is 2.2 psi.

Regulatory Guide

~

1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis I (2/76), Rev.1 Compliance 1

Comply with exception.  !

Seismic analysis of structures, systems and components in the Trojan Nuclear Plant were completed prior to the issuance of this Regulatory Guide. The methods of combining modal responses and spatial components as recommended by this Regulatory Guide are technically sound. For modifications to the Trojan Nuclear Plant and for analysis of existing structures and components, either the square root sum of the squares method or the methods of modal i '

and spatial combination recommended by this Regulatory Guide may be used at the option of the engineer performing the analysis.

However, modal responses were combined in accordance with this Regulatory Guide for the modification of the Control, Auxiliary and Fuel Building complex, with one exception.

Contrary to Position C.2 of this Regulatory Guide, the combination of effects due to three spatial components of an earthquake was not performed. Rather, a co-directional response .

Revision 8

~

TABLE 3.5-1 Sheet 16 of 35

[

v -

value was determined. That value was the maximum value obtained by adding the response due to the vertical earthquake with the larger value of co-directional response due to one of the two horizontal earthquakes by the absolute sum method.

Resrulatorv Guide 1.94 - Quality Assurance Requiremeras for Installation, Inspection, and Testing of Structural Concicite and Structural Steel During the Construction Phase of Nuclear Power Plants (4/76),

Rev.1 l

Compliance l

Comply.

Recrulatory Guide 1.102 - Flood Protection for Nuclear Power Plants (9/76), Rev.1 I l

Gmpliance O Comply with exception.

Construction of Trojan was completed pri,or to the issuance of this Regulatory Guide. The flood design criteria and bases for floods used for Trojan are described in DSAR Sections 2.4 and 3.1.4. The need for shutdown procedures and communications per Position C.2 does not apply to a permanently defueled facility that does not require operator action to maintain fuel

. integrity for several days.

Regulatorv Guide 1.105 - Instrument Setpoints (11/76), Rev.1 Compliance Comply with exception. _

Trojan does not conform to the regulatory position concerning the design verification of these instruments as part of the instrument qualification program recommended in Regulatory Guide 1.89. Trojan's setpoints meet the intent of this Regulatory Guide but not its specific Revision 8

TABLE 3.5-1 Sheet 17 of 35 format. Construction of the Trojan Nuclear Plant was completed prior to the implementation

'date of this Regulatory Guide. The method of selection of setpoints was previously approved by the NRC.

Reanlatnry Guide 1.109 - Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I (10/77), Rev.1 Comnliance Comply with exception.

(a) The operating license for the Trojan Nuclear Plant was granted prior to the issuance of this ReCulatory Guide. The methods used to estimate the preoperational doses to man from the routine operation of the Plant were reviewed and approved by the NRC.

(b) Appendix B of this Regulatory Guide provides models to be used in calculating the O annual doses to man from noble gas discharged to the atmosphere, including the dose factors to be used. PGE calculations utilize the Regulatory Guide models, but use different dose factors. The noble gas dose factors utilized in PGE calculations were reviewed and approved by the NRC.

Regulatorv Guide 1.111 - Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in

- l Routine Releases from Light-Water-Cooled Reactors (7/77), Rev.1 Comnliance Comply.

TABLE 3.5-1 Sheet 18 of 35

'Q D

Regulatory Guide 1.112 - Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors (5/77). Rev. 0-R Compliance Comply.

Regulatory Guide 1.113 - Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I (4/77), Rev.1 Compliance Comply with exception.

e Alternative models have been used as described in DSAR Section 2.4.12. These models have

( been reviewed and approved by the NRC. More specifically, dilution factors for aquatic biota, and swimming and shoreline sediment concentrations were calculated utilizing a mixing model.

The Regulatory Guide also states that in situ tracer studies can provide accurate predictions without the need of a model. Dilution factors for drinking and irrigation water were calculated by assuming complete mixing of effluent in the Columbia River. In situ tracer dye studies were used to validate this dispersion model.

Regulatory Guide 1.116 - Quality Assurance Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems (6/76), Rev. O Compliance Comply.

I O a vi > 8

4 TABLE 3.5-1 Sheet 19 of 35 O

Reguintorv Guide c 1.117 - Tornado Design Classification (4/78), Rev.1 Compliance Comply with exception.

Design, construction, and licensing activities were completed and the Trojan Nuclear Plant was operational prior to the issuance of this Regulatory Guide. All Seismic Categog I structures can withstand at least a 200-mph tornado with a tornado-induced pressure differential of 1.5 psi bursting pressure occurring in 1.5 seconds.

In addition, the following Seismic Category I structures and equipment contained therein are fully t irotected against a 300-mph tornado: i

\

Containment Centrol and cable spreading rooms 3

Auxiliary Building below grade O Both 200- and 300-mph tornadoes are greater in magnitude than any experienced in the area.

The tornado design criteria for the Trojan facility have been reviewed and approved by the NRC. ,

Regulatory Guide l

i

. 1.122 - Development of Floor Design Response Spectra of Seismic Design of Floor-Supported Equipment or Components (2/78), Rev.1 Compliance Comply with exception.

Design, construction and licensing activities were completed and the, Trojan Nuclear Plant was operational prior to the issuance of this Regulatory Guide. Design response spectra for the SSE and OBE were derived and used in seismic analyses as described in DSAR Section 3.1.1 (GDC 2). The seismic analysis for the Control, Auxiliary and Fuel Building complex was modified as described in DSAR Section 3.1.6.2.5. The methods used have been reviewed and approved by the NRC.

Revision 8

r; ,

TABLE 3.5-1 O

• Sheet 20 of 35 4 V

Regulatorv Guide j

1.123 - Quality Assurance Requirements for Control of Procurement ofItems and Services for Nuclear Power Plants (7/77), Rev.1 Withdrawn Comnliance Comply with exception.

(a) Section 6.2 of ANSI N45.2.13-1976 specifies that notification points will be used for all procurement documents. However, notification points are not applicable to all procurement activities. Notification points will be utilized by the Quality Department only when deemed necessary and appropriate.

(b) Quality-related items and services procured from a contractor, supplier, or service organization for which a quality assurance program is required but has not been reviewed or approved may be utilized by PGE with the specific written approval of the Quality Department, It is not appropriate in all cases to review or approve the quality 7 assurance program of contractors, suppliers, and service organizations. In some cases, k the supplying activity m.y not have a quality assurance program. Appropriate additional controls such as source inspection, special receipt instructions, surveillance, and testing are imposed should the supplying pr.tivity's quality assurance program not be reviewed or approved. -

Regulatorv Guide 1.140 - Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants (10/79), Rev.1 Compliance Comply with exception.

The design portions of this Reguletory Guide are not applicable to Trojan as the constmetion of the Plant was completed prior to issuance. Also, the filtration systems will no longer require in-place testing per ANSI N510-1975, since the potential offsite releases are well below.10 CFR 50, Appendix I limits without filtration. ,

TABLE 3.5-1 Sheet 21 of 35 O -

Regulatory Guide 1.143 - Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants (10/79), Rev.1 Compliance Comply with exception.

The rr.dwaste solidification system is not used at Trojan. Therefore, the guidance for solidification systems do not apply to this facility.

Reentatory Guide 1.144 - Auditing of Quality Assurance Programs for Nuclear Power Plants (9/80), Rev.1 Withdrawn Compliance Comply with exception.

(a) Section 4.3.2.4 of ANSI /ASME N45.2.12-1977 states, "When a nonconformance or quality assurance program deficiency is identified as a result of an audit, further investigation shall be conducted by the audited organization in an effort to identify the cause and effect and to determine the extent of the corrective action required." PGE investigates internal audit findings through a corrective action program that complies with PGE-8010, Nuclear Quality Assurance Program. Determination of cause may not be required if the finding does not ir.volve a significant condition adverse to quality.

(b) Section 4.5.1 of ANSI /ASME N45.2.12-1977 states, "... investigate any adverse audit findings to determine and schedule appropriate corrective action including action to prevent recurrence...." PGE does not require assignment of action to prevent recurrence if the finding does not involve a significant condition adverse to quality.

(c)

Section 4.5.1 of ANSI /ASME N45.2.12-1977 states, "In the event that corrective action cannot be completed within thirty days, the audited organization's response shall include a scheduled date for the corrective action..." PGE controls the response time for audit findings on internal programs through the corrective action program.

Evaluation completion dates are assigned based on whether the problem is a significant Revision 8 1

-J

, TABLE 3.5-1 Sheet 22 of 35

/)

V condition adverse to quality, whether an E.ER or NOV response is required, etc.

(d) Section 4.5.2.2 of ANSI /ASME N45.2.12-1977 states, " Evaluate the adequacy of the response." The auditing organization does not in all cases perform this part of the corrective action process unless the internal audit finding is a significant condition adverse to quality.

Regulatory Guide 1.145 - Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants (11/82), Rev.1 .

Compliance Comply.

Regulatory Guide Q

V 1.146 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants (8/80), Rev. 0 Withdrawn l

Compliance l Comply.

Reculatorv Guide 3.4 - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors (2/78), Rev.1-R Compliance Comply.

( ,- Revision 8

p TABLE 3.5-1 Sheet 23 of 35 d ,

Regulatorv Guide 4.1 - Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants (4/75),

Rev.1 Compliance Comply.

Reaulatorv Guide l

4.13 - Performance, Testing, and Procedural Specifications for Thermolmninescence Dos'imetry: Environmental Applications (7/77), Rev.1 .

Comnliance Comply.

'( Reguintnry Guide 4.15 - Quality Assurance for Radiological Monitoring Programs (Normal Operations) -

Effluent Streams and the Environment (2/79), Rev.1 Comoliance Comply with exception. I (a) Exception is taken for the records recommended by Section 4 of the Regulatory Guide.

Trojan will retain records of sample description, sample time, sample point, and sample results as QA Records. Calibration records for laboratory counting systems will be retained as QA records; records of " backgrounds and blanks" are not retained.

Records and methods used to control and track a sample through the sample / analysis process are of little use subsequent to sample analysis.

(b) Exception to Section 5 of the Regulatory Guide is taken to requiring specific documentation of sampler collection efficiencies. Manufacturer supplied sampler collection efficiencies are usually based on homogenous laboratory samples instead of the mixtures which are encountered in the field. Process Radiation Monitor (PRM) detector sensitivities and calibration requirements are specified in the Offsite Dose Revision 8 2

f TABLE 3.5-1 Sheet 24 of 35 Calculation Manual (PGE-1021). Gaseous grab sample results are compared to related PRM readings to provide confidence in the accuracy of the grab sample. Liquid grab sample collection efficiency (representative sample) is ensured by recirculating two tank volumes and purging sample lines prior to drawing the sample. Trojan additionally takes exception to " periodically take replicate grab samples to determine reproducibility of sampling". Grab sampling will be performed as required.

(c) Exception is taken to routinely analyzing " replicate" samples and blank and spiked samples to be submitted for analysis as unknowns, per Section 6.3.1 of the Regulatory Guide. Replicate, blank, and spiked samples are not run routinely; known analytical blank samples are analyzed by each technician as a part of the chemistry training program specified in procedures.

(d) Exception is taken to Section 7 of the Regulatory Guide. ' Trojan will comply with the ,

requirements for calibration and checks of effluent monitoring systems which are  !

contained in the Offsite Dose Calculation Manual (PGE-1021).

(e) Exception is taken to Section 9 of the Regulatory Guide for auditors being " qualified" in radiochemistry. Records of backgrounds and blanks are not necessary in this O- circumstance. Audits of this area will be performed by knowledgeable personnel.

Repuistory Guide 5.7 - Entry / Exit Control for Protected Areas, Vital Areas, and Material Access Areas (5/80), Rev.1 Compliance Comply with exception.

The Trojan security program was established and operated in accordance with the " Trojan Nuclear Plant Security Plan", PGE-1017, and Regulatory Guide 5.66, which applies directly to operating nuclear power plants. After permanent defueling of the facility the security program was revised and implemented in accordance with the " Trojan Nuclear Plant Security Plan", PGE-1017. Access controls in the NRC approved Trojan Security Plan to prevent entry of unauthorized personnel, vehicles or materials into restricted areas closely parallel and do meet the intent of this Regulatory Guide. The criteria in this Regulatory Guide are intended to apply to a broader category of nuclear facilities than just nuclear power plants and it is therefore not directly applicable.

l Revision 8 i

TABLE 3.5-1 O Sheet 25 of 35 Regulatory Guide 5.12 - General Use of Locks in the Protection and Control of Facilities and Special Nuclear Materials (11/73), Rev. C Comnliance Comply with exception.

PGE has complied with the provisions of this Regulatory Guide in the use of locks intended for physical protection purposes at the Trojan Nuclear Plant only to the extent described in the

" Trojan Nuclear Plant Security Plan," PGE-1017. The Plan is an NRC approved document which contains Safeguards Information and the details are withheld from public disclosure.

After permanent defueling of the facility the security program was revised and implemented in accordance with the " Trojan Nuclear Plant Security Plan," PGE-1017.

i Regulatorv Guide A 5.14 - Use of Observation (Visual Surveillance) Techniques in Material Access Areas (5/80),

V Rev.1 1

Compliance Comply with exception.

The criteria delineated in this Regulatory Guide are primarily applicable to processing facilities in which special nuclear material may be unalloyed or unencapsulated and thus can be susceptible to unauthorized diversions. Special nuclear material (SNM) at Trojan is in the form oflarge fuel assemblies which are not concealable or susceptible to theft, and the visual surveillance techniques of this Regulatory Guide are not necessary for its control.

l l

l V- Revision 8 i

i

I I

s TABLE 3.5-1 Sheet 26 of 35 Reentatory Guide 5.20 - Training, Equipping, and Qualifying of Guards and Watchmen (1/74), Rev. O Compliance Comply with exceptions.

10 CFR 73, Appendix B, specifies in much greater detail than RG 5.20 the exact security force training and qualification requirements. Report PGE-1024, " Trojan Nuclear Plant Security Force Training and Qualification Plan," an NRC approved document, was drafted to comply with the requirements of 10 CFR 73, Appendix B, and~the guidelines of NUREG-674,

" Security Personnel Training and Qualification Criteria," in lieu of RG 5.20 with the following limited exception: security officers wear uniforms which are markedly distinct from those of local law enforcement authorities (reference: RG 5.20, Section 4.0).

Regulatory Guide 5.29 - Nuclear Material Control Systems for Nuclear Power Plants (6/75), Rev.1 Compliance ,

Comply with exception. ,

This Regulatory Guide endorses ANSI N15.8-1974 as being acceptable to the NRC to provide an adequate basis for the control and accounting of SNM at nuclear power plants. The following exceptions are taken to ANSI N15.8-1974:

(a) The nuclear material manager is not independent of the Plant staff as recommended by Paragraph 3.1.1 The SNM custodian performs the duties of both the nuclear materials j manager and the nuclear materials custodian. SNM documentation is checked by I Quality Assurance. Since Trojan is the only nuclear facility owned by PGE. a separate company nuclear materials manager is not needed.

(b) Audits of the fuel suppliers are not conducted by PGE as recommended by Paragraph 5.1. An audit by PGE of the fuel supplier is unnecessary since the supplier must have an SNM license and, as part of maintaining that license, must have a j material control system. The SNM license holder is already subject to periodic NRC  !

and DOE audits. PGE verified that the supplier has a valid SNM license. l Revision 8 I

TABLE 3.5-1 O Sheet 27 of 35 l

(c) Physical inventories of irradiated fuel are'done on a 12-month interval, instead of a 6-month interval as recommended by Paragraphs 6.4.1 md 6.4.2. 10 CFR 70.51(d) requires that physical inventories be conducted at inmcvals no greater than 12 months i and is the basis for PGE's 12-month interval. l (d) Irradiated fuel inventories are done by item count and do not explicitly include serial number identification as recommended by Paragraph 6.4.2. Special viewing equipment is necessary for serial number identification due to difficulty in seeing the numbers.

Reliance is placed on administrative controls for fuel positioning and item count for quantity verification. ,

(e) Paragraph 6.4.3 concerning serial number verification [ luring unload is no longer applicable. Trojan is permanently defueled and no fuel will be placed in the reactor vessel.

(f) Audits of reprocessors are not conducted by PGE as recommended by Paragraph 8.2.

An audit by PGE on the reprocessor is unnecessary for the same reasons as given in (b) above.

s (g) i Item control areas are not classified as vital areas as recommended by Paragraph 11.  !

No vital equipment or vital areas exist at the Plant in the defueled configuration.

However, access control, surveillance, and other physical security measures exist which are applicable to item control areas. These are described in PGE-1017.

Regulatory Guide I j

)

5.43 - Plant Security Force Duties (1/75), Rev. I l

Compliance l

Comply with exception.

The " Trojan Nuclear Plant Security Plan," PGE-1017, an NRC approved document, describes the organization and duties of the security force. This Regulatory Guide has been considered in the development of the Trojan security force; however, PGE does not fully comply with it.

For further information, refer to the Trojan Security Plan since these details are considered to be Safeguards Information and are withheld from public disclosure. After permanent defueling of the facility the security program was revised and implemented in accordance with the " Trojan Nuclear Plant Security Plan," PGE-1017. '

Revision 8

-i .

n.

TABLE 3.5-1 O, '

Sheet 28 of 35 Regulatory Guide 5.44 - Perimeter Intrusion Alarm Systems (5/80), Rev. 2 Comnliance Comply with exception.

The " Trojan Nuclear Plant Security Plan," PGE-1017, an NRC approved document, describes the Trojan Protected Area intrusion alarm system for an operating facility. Although the system meets much of the criteria of Revision 2 to this Regulatory Guide, specific commitments only to comply with certain portions of Revision 1 of this Regulatory Guide have been made. With the NRC approval of the system, additional modi 0 cations are not considered necessary. For further information, refer to the Trojan Security Plan since these details are considered to be Safeguards Information and are withheld from public disclosure.

Subsequently, based on the defueled condition of the Plant, isolation zones and perimeter intrusion detection systems are no longer necessary at the facility. The security program was

(-

\

revised and implemented in accordance with the " Trojan Nuclear Plant Security Plan,"

PGE-1017.

Regulatory Guide 5.54 - Standard Format and Content of Safeguards Contingency Plans for Nuclear Power Plants (2/89), Rev. 0 Compliance Comply with exception.

Trojan's Safeguards Contingency Plan is incorporated into the Trojan Security Plan. Although the format and content differ slightly from the guidance given in this Regulatory Guide, the content and format of Trojan's Safeguards Contingency Plan are acceptable to the NRC as indicated by approval of the Trojan Security Plan.

Revision 8 1

TABLE 3.5-1 Sheet 29 of 35 V

Regulatory Guide 5.57 - Shipping and Receiving Control of Special Nuclear Material (6/80), Rev.1 Comnliance Comply with exception.

This Regulatory Guide is primarily intended for nuclear fuel fabricators and reprocessors in which theft of small quantities of SNM is possible, and as such, is only loosely related to nuclear power reactors. The NRC has provided specific guidance to nuclear power plants in Regulatory Guide 5.29, and PGE has complied with those provisions as clarified in the discussion for that Regulatory Guide.

Regulatorv Guide 5.62 - Reporting of Safeguards Events (11/87), Rev.1 Compliance Comply with exception.  !

The NRC amended 10 CFR 73.71(c)(2) and 10 CFR 73, Appendix G (60 FR 13615, dated {

March 14,1995), effective April 13, 1995. These amendments eliminate the requirement for quarterly submittal of safeguards event logs. Therefore, the Plant complies with this Regulatory Guide with the exception of those provisions specifically related to quarterly

. safeguards event log submittal.

Regulatory Guide 5.65 - Vital Area Access Controls, Protection of Physical Security Equipment, and Key and Lock Controls (9/86), Rev. O Comoliance Comply with exception. t During plant operation the facility security program, which was approved by the NRC, was implemented in accordance with the " Trojan Nuclear Plant Security Plan," PGE-1017.

l O

'y) Revision 8 l

[

TABLE 3.5-1 Sheet 30 of 35 U

After permanent defueling of the facility the security program was revised and implemented in accordance with the " Trojan Nuclear Plant Security Plan," PGE-1017, which includes the following exceptions.

Positions C.1 to C.4 apply to Vital Areas and are no longer applicable at Trojan as no vital equipment nor vital areas exist at the Plant in the defueled configuration.

q Position C.5 is concerned with suspending security measures. The Plant complies with the 4

exception that Certified Fuel Handlers and not SROs approve the suspension.

Position C.6 deals with the periodic review of the Security and, Contingency Plans to determine their effect on safety and the Plant complies.

Position C.7 requires security onsite secondary power supplies to be in vital areas. PGE has requested an exception from this requirement.

Position C.8 is concerned with security keys and locks and the Plant complies.

p Regulatory Guide V

5.66 - Access Authorization Program for Nuclear Power Plants (6/91), Rev. O Compliance .

Comply.

. Reaulatory Guide 7.1 -Administrative Guide for Packaging and Transporting Radioactive Material (6/74), Rev. O Comnliance Comply.

V Revision 8

TABLE 3.5-1 Sheet 31 of 35 Reculatory Guide 7.2 - Packaging and Transportation of Radioactively Contaminated Biological Materials (6/74),

Rev.O Compliance Comply.

4 Regulatorv Guide 7.4 - Leakage Tests on Packages for Shipment of Radioactive biaterials (6/75), Rev. O Compliance Comply.

Regulatorv Guide

(

7.7 - Administrative Guide for Verifying Compliance with Packaging Requirements for Shipments of Radioactive Materials (8/77), Rev. O Complisnce

• Comply.

Reculatory Guide 8.1 - Radiation Symbol (2/73), Rev. O Compliance Comply.

rh V Revision 8

TABLE 3.5-1 Sheet 32 of 35 O

Regulatory Guide 8.2 - Guide for Administrative Practices in Radiation Monitoring (2/73), Rev. O Compliance ,

Comply.

Regulatorv Guide 8.4 - Direct-Reading and Indirect-Reading Pocket Dosimeters (2/73), Rev. O Compliance Comply with exception.

Pocket dosimeters are calibrated to il5 percent, in lieu of the i10 percent value in the Regulatory Guide. TLD's are the primary method of personnel radiation exposure O monitoring. Pocket dosimeters are used for initial estimations and backup dosimetry. Plus or minus 15 percent is not significantly different than i10 percent as per the Regulatory Guide.

Reculatory Guide l

8.7 - Occupational Radiation Exposure Records Systems (5/73), Rev. O Compliance Comply.

Regulatorv Guide 8.8 - Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable (6/78), Rev. 3 Comoliance Comply with exception.

Position C.2 and parts of Position C.4 of the Regulatory Guide relate to construction of new -

Revision 8 l

\

O ,

TABLE 3.5-1 ,

Sheet 33 of 35 plants. Section D states that the guidance is to be used during the construction permit and operating license stages of plant licensing. Section D also states that, for plants desigred before the guidance was issued (such as Trojan), no substantive design changes will be required at the operating license stage unless the design change can prevent substantial man-rem exposures that cannot be prevented by procedural measures, and the design change is consistent with the cost effectiveness principle of maintaining occupational radiation exposures ALARA. PGE will comply with Positions C.I.d, C.2 and C.4 on a case-by-case basis when Plant design modifications are made.

Reaninforv Guide 8.9 - Acceptable Concepts, Models, Equations, and Assumptioits for a Bioassay Program (9/73), Rev. 0 ,

Comnliance Comply.

Reculatory Guide 8.10 - Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably . Achievable (5/77), Rev.1-R Compliance Comply.

Regninforv Guide 8.13 - Instruction Concerning Prenatal Radiation Exposure (12/87), Rev. 2 Compliance Comply.

O Revision 8

)

_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _____ i

l TABLE 3.5-1 Sheet 34 of 35 Regulatory Guide 8.15 - Acceptable Programs for Respiratory Protection (10/76), Rev. O Compliance Comply with exception.

This Regulatory Guide cites NUREG-0041 as containing detailed advice and background information on an acceptable program. NUREG-0041 will be used for guidance and background information only, and will not be complied with fully (verbatim). Other ways different from NUREG-0041 may be more appropriate for a particular program element.

NUREG-0041 has not been updated or revised since its original October,1976 publication. It contains some errors, inconsistent statements, and out-of-date information, for which full compliance is not appropriate.

Industry standard ANSI Z88.2, which is the basis for this Regulatory Guide, has been extensively revised. Section 6.5 of this revised standard now permits the use of contact lenses Q with respirators. The NRC, in a memorandum dated June 5,1989, and the Occupational Safety and Health Administration, in a memorandum dated February 8,1988, have changed their regulatory positions to permit the use of contact lenses with respirators.

Reculatory Guide 8.25 - Calibration and Error Limits of Air Sampling Instruments for Total Volume of Air

. Sampled (8/80), Rev. O Compliance Rev. 0 - Comply.

Regulatory Guide 8.26 - Applications of Bioassay for Fission and Activation Products (9/80), Rev. O Compliance Comply.

A V Revision 8

p

3. .

l- .

L-TABLE 3.5-1 Sheet 35 of 35 Reaulatnry Guide I

8.27 - Radiation Protection Training for Personnel at Light-Water-Cooled Nuclear Power Plants (3/81), Rev. O Comnliance

{

l Comply with exception.  !

(a) l Position C.1 states that trainers must have a knowledge level exceeding that of the 1 workers completing the training. PGE's position on this guidance is that instmetors j need only be qualified to instruct within the scope of what they are teaching. l Otherwise, a situation could develop where a visiting certified health physicist could I not be trained because his knowledge level in certain as~pects of the training program L could exceed that of the training staff.

(b) Position C.1 recommends training times for various categories of workers, varying from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for non-radiation workers to 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> for closely supervised workers to

's 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for those workers operating independently. PGE's position is that workers

]

will receive training, as per 10 CFR 19.12, commensurate with the potential radiation j protection problems in their work area. Workers employed in tasks beyond general employee training will receive specific additional training under the plant ALARA program. No specific requirement's for amount of training time are necessary.

t (c) Section 3.2.6 recommends mock-up training for jobs when the collective dose exceeds j 1 man-rem. PGE's ALARA program requires pre-job reviews when the collective '

dose exceeds 2 man rem. PGE's position is that the graduated review program as  ;

exists in the Plant's ALARA program is sufficient to meet the requirements of 10 CFR '

19.12 and 10 CFR 20. j Renulatorv Guide

8.29 - Instruction Concerning Risks from Occupational Radiation Exposures (7/81),

Rev.O Compliance Comply.

j .r .

useemis Revision 8

z Corrective measures for out of specification chemistry will be taken upon discovery. The Modular SFP Cooling and Cleanup System is used to maintain satisfactory SFP purity and clarity. The capability for boric acid addition is required to ensure SFP coolant can be mainutined >_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 instnunentation 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 discharge pressure, temperatures throughout the system, and radiation level from the air cooler enclosure drain line. Alarms are provided to the control room for abnormal SFP level or high SFP temperature. The radiation detector monitors for leakage from the air coolers or piping inside the enclosure and automatically shuts down the system if leakage is detected. There are no other 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 Instrumehtation The seismic ' monitoring instmmentation, consisting of a multi-element seismoscope and peak acceleration recorders, is subject to routine maintenance and testing to ensure it is available to record data from seismic events.

SR-6341 Multi-element Seismoscope SR-6340B Peak Acceleration R.ecorder - Intake Structure,45-foot Elevation l SR-6340D Peak Acceleration Recorder - Control Building Mezzanine, Top of Ladder Above Secondary Sample Storage Room SR-6340E Peak Acceleration Recorder - Fuel Building, 93-foot Elevation Hot Shop, West, Behind the Wall 4.1-3 Revision 8 t

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

I 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.

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, frequency 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 and components that support the emergency plan, fire protection plan, security plan or other licensed condition.

O e

Revision 8 4.1-4

' 4.2 SPENT FUEL HANDLING 4.2.1 SPENT FUFI RECFTPT. HANDLING. AND TRANSFER l

Movement of new or spent fuel into the Containment Building is not authorized without prior l NRC approval. Moving spent fuel assemblies within the SFP and loading spent fuel and other l

- materials into Transfer and Storage Casks in the Fuel Building is authorized. SFP fuel l handling activities are accomplished by use of the Spent Fuel Pool bridge crane and spent fuel l.

handling tool. However, loading spent fuel into Transfer and Storage Casks requires l additional equipment and functional considerations not specifically described in this section. l .

~

' Additional details regarding special equipment and procedures for transferring spent fuel l between the SFP and Transfer and Storage Casks ae provided in the Trojan ISFSI SAR, l LCA 237, and LCA 246.

l l

For fuel movement within the SFP, inadvertent criticality prevention is provided as descHbed l in Section 4.1.1. Criticality prevention associated with loading spent fuel into Transfer and l l'

Storage Casks is as described in the ISFSI SAR, LCA 237, and LCA 246. -

l 4

4.2.1.1 Functional Description Figure 4.2-1 provides a layout of the spent fuel storage area. Movement of spent fuel assemblies is accomplished by use of the spent fuel pool bridge crane and spent fuel handling tool.

The SFP bridge crane consists of a wheel mounted walkway spanning the SFP. The SFP bridge crane moves in the north-south direction on rails by means of a two-speed motor. The SFP hoist is a 1-ton capacity electrical monorail hoist that moves in the east-west direction on the overhead structure of the SFP bridge crane. The fuel assemblies are moved within the SFP by means of a long-handled tool suspended from the SFP hoist. The spent fuel handling tool- ,

4.2-1 Revision 8

is a manually actuated tool used to handle spent fuel in the SFP. Fuel assembly inserts, such as thimble plugs, burnable poison rods, rod control clusters, and source rods, may also be transferred between positions within the SFP.

Load movements outside of the SFP may be provided by the Fuel Building crane. The Fuel Building crane is a 125-ton pendant-operated crane and is provided with an auxiliary hoist rated at 25 tons. The Fuel Building crane is restricted from moving loads over the SFP. The Fuel Building crane is also limited to moving in a way that avoids the possibility of falling or dropping objects into the SFP. Mechanical stops installed on the rails prevent the crane hook l from traveling closer than six feet from the SFP. Electrical limit switches deenergize the bridge drive before the mechanical stops. These stops and limit switches may only be bypassed or removed while following an approved procedure, and the' n only with the Shift Manager's approval.

4.2.1.2 Safety Features The SFP bridge crane incorporates design features to minimize the probability of a fuel handling accident. These features are discussed in Section 3.3.1.

Adminis:rative controls in place to minimize the probability and consequences of fuel handling accidents include:

(1) Fuel handling operations can only be performed under the direct supervision of a CFH.

l I

(2) Fuel handling operations must be performed in accordance with approved Plant procedures.

Revision 8 4.2-2 O

1

(3) Loads carried over the SFP and the heights at which they may be carried over I Q storage rackh containing fuel shall be limited to preclude impact energies over 240,000 in-lbs.

l 4.2.2 SPENT FUEL STORAGE Section 3.2.2 provides a description of the SFP design including the storage racks. Although SFP storage rack capacity is 1287 fuel assemblies, only 781 fuel assemblies are stored.

The SFP Cooling System provides forced cooling of the spent fuel assemblies. Bulk SFP coolant temperature is maintained between 40*F and 140 F. A more thorough description of 1

the operation of the SFP Cooling System is provided in Section 4.3.1,. j i

l Subcritical arrays are ensured by maintaining center-to-center distance between adjacent fuel assemblies and the fixed neutron absorber contained in the spent fuel storage racks.

Additional discussion is provided in Section 4.1.1.

1 Radiation shielding is provided by the water level maintained in the SFP. A minimum of 23 feet of water is maintained above the top of a spent fuel assembly in the storage racks.

During fuel transfer operations, at least 9.5 feet of water is maintained above the top of the j

. active portion of a fuel assembly. This water barrier serves as a radiation shield enabling the i gamma dose rate at the pool surface from the spent assembly to be maintained at or below 2.5 mrem /hr.

4.2-3 Revision 8

4.5 REFERENCES

( .

REFERENCES FOR SECTION 4.1

1. Trojan Nuclear Plant Finni Safety Annivsis Reoort, Amendment 19 (December 1992). l

2. _PGE-1012. Trojan Nuclear Plant Fire Protection Program.

l REFERENCES FOR SECTION 4.2

1. Troian Ooeratina Iicence. No. NPF-1, Amendment 200'(April 23,1999). l

2. _ Trojan Nuclear Plant Final Safety Annivsis Reoort, Amendment 19 (December 1992). l l l

, 3. PGE-1037. Trojan Nuclear Plant Spent Fuel Storage Rack Replacement Reoort. l Historical only.

O 4. PGE Calculation. TC-720. Trojan Spent Fuel Pool Heatup and Boil Down Races, Revision 4 (November 12,1996).

l

5. PGE-1069. Troian Indeoendent Soent Fuel Storace Installation Safety Annivsis Reoort.

l l

6. License Chance Anolication (LCA) 237. Soent Fuel Cask Loadino in the Fuel Buildino. l Transmitted by PGE letter, VPN-005-99, dated January 7,1999. l l

7. LCA 246. Soent Fuel Cask Loading in the r uel Building - Contingency Fuel Unloading l to the Fuel Pool. Transmitted by PGE letter, VPN-006-99, dated January 27,1999. l REFERENCES FOR SECTION 4.3

1. Trojan Nuclear Plant Final Safety Analysis Report, Amendment 19 (December 1992). l

2. PGE-1012. Trojan Nuclear Plant Fire Protection Plan. l 4.5-1 Revision 8

i 510FFGAS TREATMENT AND VENTILATION The Offgas Treatment and Ventilation Systems consist of the Containment Purge Exhaust and l Fuel and Auxiliary Building Ventilation Systems, which have the potential for discharge of radioactive effluents.

5.2.1 Deleted j 5.2.1.1 Deleted l 5.2.1.2 Deleted l 5.2.1.3 De:eted l 5.2.2 CONTAINMENT VENTILATION SYSTEM The Containment Ventilation Systems, depicted in Figures 5.2-2 and 5.2-3, provide a means of ventilating the Contaimnent Building. The, systems provide for removal of potentially contaminated gases from within the Containment Building and exhausts them to the environs. The systems involved are containment purge supply (CS-1) and containment purge exhaust (CS-2).

5.2.2.1 Design Bases The design bases for the Containment Ventilation System are as follows:

(1) Control containment airborne activity levels in order to allow containment access while limiting personnel exposure to less than the dose limits for occupational exposure of 10 CFR 20.

(Q/

5.2-1 Revision 8

(2) Control and filter the releases of gaseous effluents during coritainment purge such that the intent of 10 CFR 50, Appendix I, may be met for overall Plant radioactivity 3 releases.

I 12.2.2 System Description The Containment Purge Supply System (CS-1) consists of the following equipment for operation in the defueled condition:

(1) An outside air intake (2) A prefilter (3) A bank of HEPA filters (4) Two purge supply fans (5) Backdraft dampers (6) Ductwork, valves, and instrumentation The fans and filters are located outside the containment.

The Containment Purge Exhaust System (CS-2) censists of a the following equipment:

(1) Prefilters (2) A bank of HEPA filters 1

Revision 8 5.2-2 O

l l)

(3) Two purge exhaust fans f

Q/. '

(4) Backdraft dampers 4

(5) Ductwork, valves, and instrumentation

, The fans and filters are located outside the containment. The system exhausts via the containment purge vent at the top of the Containment Building. The exhaust air is monitored for radioactivity by PRM-1, the Containment Monitoring System. See Section 5.5 for further discussion of PRM-1.

The containment ventilation systems are used as necessary to support containment access.

R2.3 Design Evaluation The containment purge and exhaust systems provide no safety functions. They are used as necessary to support containment access.

l 5.2.3 FUEL AND AUXII TARY BUILDING VENTILATION SYSTEM The Fuel and. Auxiliary Building Ventilation Systems, depicted in Figures 5.2-4, 5.2-5, and 5.2-6, provide a means of ventilating the Fuel and Auxiliary Buildings and maintains a negative atmosphere in spaces subject to airborne radioactive contamination. The systems involved include the Fuel and Auxiliary Building Supply System (AB-2), the Fuel and. Auxiliary Building Exhaust System (AB-3), and the Spent Fuel Pool (SFP) Exhaust System (AB-4).

r

-(

5.2-3 Revision 8

5.2.3.1 Desien Bases O

'(1) AB-2 is designed to supply outside ventilation air to the Fuel and Auxiliary Buildings. Design objectives applicable to the operating plant condition for building temperature control are not applicable for the permanently defueled condition.

(2) AB-3 and AB-4 are designed to filter exhaust air before release to the environs, providing high efficiency removal of particulates.

5.2.3.2 System Description The Fuel and Auxiliary Building Ventilation Systems consist of the following equipment:

(1) Two AB-2 supply fans (2) Four AB-3 exhaust fans h

1 (3) Two AB-4 exhaust fans ,

(4) Roll filters (5) HEPA filters in AB-3 and AB-4 (6) Backdraft dampers (7) Ductwork, valves, dampers, and instrumentation At full capacity, the Ventilation System is designed to provide approximately six air changes per hour within the buildings. -

O Revision 8 5.2-4 l

f

The exhaust from the Fuel and Auxiliary Building Ventilation Systems is monitored by the Auxiliary Building Vent Exhaust Monitoring System (PRM-2). PRM-3 monitors the AB-4 discharge from the SFP area into the Fuel and Auxiliary Building ventilation exhaust duct. See Section 5.5 for further discussion of PRM-2 and PRM-3.

5.2.3.3 Design Evahintion The Fuel and Auxiliary Building Ventilation systems provide no safety functions.

AB-3 and AB-4 limit the spread or release of airborne radioactive material by filtering the ventilation exhaust prior to discharge to the environs.

At full capacity, the Ventilation System is designed to provide approximately six air changes per hour within the buildings. Full capacity operation of the Ventilation Systems is not expected to be normally required for the permanently defueled condition due to the limited amount of work anticipated to be performed and the decreased amount of radioactive O material present.

U l

The HEPA filters installed in the exhaust systems will provide for cleanup of airborne radioactive I material within the buildings and minimize the release of radioactive particulates to the I environment. I l

5.2.4 RADWASTE PROCESSING BUH DING VENTILATION SYSTEM The Radwaste Processing Building Ventilation System provides a means of ventilating the Radwaste Processing Building and maintaining a negative atmosphere in spaces subject to airborne radioactive contamination. Radwaste storage, handling, and processing operations take place in the.

building. Portions of the Ventilation System have been deactivated.

5.2-5 Revision 8

S.2.4.1 Design Bases The Radwaste Processing Building Ventilation System is designed to filter exhaust air before release to the environs, providing high efficiency removal of particulates. Supply air is provided through infiltration and, as appropriate, through use of the roof supply fans.

5.2.4.2 System Descriotion The Radwaste Processing Building Ventilation System consists of the following equipment:

(1) Two exhaust fans -

(2) Supply fans (3) HEPA filtration (4) Ductwork, instrumentation, and sample taps The ventilation exhaust ducting contains sample connections that are used with a fixed-filter assembly for collection of air particulates. The sampling arrangement consists of a sample probe g

downstream of the HEPA filter, connected through valves and tubing to a filter sampler, an in-line rotameter, and an air pump. An averaged' rate'of sample flow will be used in the sample analysis, the frequency of which will be periodically adjusted based on previous analysis results and the type of work currently in progress.

Building supply fans are operated as appropriate for workspace habitability. Operation of the fans is controlled to maintain a negative pressure in the building during radwaste processing activities.

Routine air monitoring is performed during radwaste processing activities in the building.

Revision 8 5.2-6

)

b

Details on sample and analysis schedules are described in the Trojan Nuclear Plant Offsite Dose Calculation Manual. The schedules comply with the NRC positions described in Regulatory Guide 1.21.

L 5.2.4.3 Design Evaluntion The Radwaste Processing Building ventilation System provides no safety function. The system limits the spread or release of al. borne radioactive material by filtering the ventilation exhaust prior to discharge to the environs.

' The system is operated and maintains a negative atmosphere while the building is occupied. The HEPA filter installed in the exhaust duct provides for cleanup of airborne radioactive material within the building and minimizes the release of radioactive particulates to the environment.

5.2.5 MODULAR SPENT FUFT POOL COOT.ING SYSTEM COOLING AIR The Modular SFP Cooling System uses outside cooling air to cool the SFP. This system is not considered a building ventilation system since it provides no building habitability functions.

The system design bases are described in Section 3.3.2.

As noted in Section 5.2.3, the Fuel and Auxiliary Building Ventilation System maintains a negative atmosphere in spaces subject to airborne radioactive contamination. When both Modular SFP Cooling System cooling fans are running, the portion of the air cooler enclosure upstream of the fans operates at a slight negative pressure relative to the Fuel Building. This is acceptable since the enclosure is separate from the Fuel Building environment. There is no mixing.of the outside cooling air with the air inside the Fuel Building and the enclosure does not provide a release path for airborne radioactive material that might be present inside the Fuel Building adjacent to the enclosure.

(

5.2-7 Revision 8

)

l l

5.3 LIOUID WASTE TREATMENT AND RETENTION O .

The Liquid Waste Treatment and Retention System (LRWS), shown in Figure 5.3-1, typically collects, stores, processes, monitors, and discharges plant liquid effluents that may contain radioactive nuclides. The system consists of components from the former Clean Radioactive Waste Treatment System (CRWS) and the Dirty Radioactive Waste Treatment System (DRWS).

The CRWS and DRWS have been consolidated by plant modification to form the LRWS. Where other licensing documents refer specifically to the CRWS or DRWS, these terms are now interchangeable with the LRWS.

5.3.1 Desicm hees .

The following design bases apply to the LRWS:

1 1

(1) To provide a means for collecting liquid effluent from floor and equipment drains l  !

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

supported systems ll (3) To reduce the concentration of nuclides and particulates in the stored liquid to a level

, that will permit release to unrestricted areas within the guidelines of Appendix I to 10 CFR 50, and limit the dose to any organ to < 5 mR during any calendar yearm (4) .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 (5) To measure and record the release of plant liquid radioactive effluent O 5.3-1 Revision 8 l

l S.3.2 System Description The typical sources of liquids collected in the LRWS are various floor and equipment drams, washdown water from decontamination evolutions, and flush water. Processed effluent from the LRWS is discharged.

The system consists of the following components:

(1) Two Treated Waste Monitor Tanks (TWMT)

(2) Two Treated Waste Monitor Tank pumps -

(3) Dirty Waste Drain Tank (DWDT)

(4) Two Dirty Waste Drain Tank pumps (one electrically deactivated) l (5) Two influent bag filters O

l (6) Two effluent bag filters

, l (7) Liquid radwaste discharge process and effluent radiation monitor (PRM-9) l (8) Interconnecting piping, valves, and instrumentation Liquids processed by the LRWS may contain varying amounts of boric acid and other chemicals.

Piping, valves, and major components (with the exception of certain radwaste pump strainers) in contac: with process fluids are constructed of corrosion resistant materials.

l System tanks are vented. Tank overpressure protection is provided by tank overflows to the l floor. .

Revision 8 5.3-2

i The DWDT is divided into two equal sections. The DWDT pump can be aligned to either tank l section. The pump is typically used to transfer tank contents through the influent bag filters to ,

i the treated waste monitor tanks.

The influent bag filters are normally used to remove particulate matter discharged from the

. system collection tanks. Differential pressure indication is provided to alert the operator to excessive filter clogging that will require filter replacement.

The TWMTs receive fluids from the system collection tanks and store the fluids for monitoring and possible further processing prior to discharge. The TWMT pumps can be aligned to either tank.' The pumps are typically used to recirculate the tank contents for sampling prior to discharge, transfer the waste from tank to tank through the effluent bag filters (and/or optional demineralizer) for additional processing, or to discharge the tank contents. Mixing eductors on recirculation lines, internal to the tanks, are designed to achieve thorough mixing to ensure l representative sampling.

l The effluent bag filters are normally used to remove particulate matter from the TWMT contents, if additional processing is determined to be required prior to discharge based upon sample results.

Differential pressure indication is provided to alert the operator to excessive filter clogging that will require filter replacement. Additionally, connections exist for an optional demineralizer if

. filtration is not adequate to support discharge requirements.

The plant discharge header receives liquid waste from the TWMTs and discharges the waste to the Columbia River by way of the discharge and dilution structure. The plant discharge header flow signal is used to regulate plant discharge flow by controlling the header flow control valve, which maintains the discharge rate at a preset value. PRM-9 monitors the discharge header for radioactivity and actuates the header isolation valve to stop discharge flow if excessive radioactivity is detected. Section 5.5 provides further discussion of PRM operation. In addition, discharge flow will be stopped whenever insufficient dilution flow is detected at the discharge and ,

N 5.3-3 Revision 8

4 dilution structure. Flush water for header cleanup following a discharge may be provided from varietig non-contaminated sources.

The discharge and dilution structure is discussed in Section 3.3.9.

5.3.3 Design Evaluation The LRWS 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. Ihe LRWS has adequate capacitp to process waste from the small number of input sources in op. ration with the plant in the permanently defueled condition.

Administrative controls are established to require further processing when recessary to ensure efnuent releases are within allowable limits.

I O

l i

O Revision 8 5.3-4

I 5.4 SOLID WASTES

• The Solid Radioactive Waste System (SRWS) provides for storage and processing for disposal of

~

radioactive solid wastes generated at the Plant. These wastes include spent demineralizer resin, expended filters, and miscellaneous contaminated equipment and solid refuse.

5.4.1 DESIGN BASES The design bases of the SRWS include the following:

1 (1) To provide a means of packaging spent demineralizer resins and expended filters in l disposable containers suitable for transfer from the Plant site (2) To provide adequate shielding in storage areas for retaining wastes in disposable l containers pending shipment to appropriate disposal facilities O

(3) To provide a means of packaging miscellaneous contaminated solid wastes generated by l sampling, ventilation filter replacement, decontamination refuse and various other items resulting from maintenance  ;

(4) To utilize shipping containers and procedures which conform with the regulations of l Title 10, Parts 20 and 71, and Title 49 of the Code of Federal Regulations (5) To observe, measure, and record the contamination levels of solid radioactive wastes l which are processed for shipment from the site O

5.4-1 Revision 8 e

5.4.2 SYSTEM DESCRIPTION O

Inputs to the SRWS include the following:

(1) Spent demineralizer resins which have been used to process potentially radioactive liquid (2) Expended filters which have been used to process potentially radioactive liquid (3) Miscellaneous solid wastes which are potentially contaminated

~

5.4.2.1 Soent Resin Transfer Sysic;n Spent demineralizer resins are transferred directly to a disposable liner. Water is removed from the disposable liner and returned to the Liquid Radioactive Waste System.

5A.2.2 Filter Handling In general, potentially contaminated filters will require replacement when filter clogging causes excessive differential pressure across the filter or when radiation levels from filter sludges become excessive.

1 l Filters are manually removed from filter housings for disposal.

5.4.2.3 Solid Wastes Dry, solid radwaste processing is typically performed in the Radwaste Annex, Containment, free l l

release facility, or in the Radwaste Processing Building (formerly the Condensate Deminerahzer

]

Building). Some segregation of radwaste material takes place within the plant buildings. The j i'

Radwaste Processing Building is typically used for cutting, decontamination, and/or packaging, solid radioactive waste.

Revision 8 5.4-2 L... . _ _ _ _ _ _ - - _ - - _ - _ - _ _ l

The Radwaste Annex to the Fuel Building includes a drum compactor, which is used for O

V compacting dry, active wastes. A free release facility (formerly the Emergency Diesel Generator rooms) is used to perform final surveys of decontaminated solid radwaste intended for free

^

release. A solid waste compactor may be used to compact miscellaneous solid waste materials into drums for storage and shipment offsite.

5.4.3 DESIGN EVALUATION I The SRWS does not perform any safety functions. The volume of spent resin required to be 1

(

processed with the plant permanently defueled is significantly less than with the plant operating.

~

In addition, the maximum expected activity of the spent resin volume is expected to be far less 4 than with the plant operating. The maximum expected activity of the spent resin volume was conservatively based on the resin fission product activities for plant operation with reactor coolant activity levels determined on the basis of fission product diffusion through cladding defects in 1.0 percent of the fuel rods.

Similarly, volume and maximum expected activity associated with expended fiiters and miscellaneous solid wastes for the permane,ntly defueled plant condition are bounded by the analysis for plant operation.

l O  !

5.4-3 Revision 8 1

www-

. Each gaseous or liquid monitoring system is provided with a sample chamber that is sized and shielded as required to ach'ieve the required sensitivities. Each sampler is constructed to instrument design standards consistent with their service and the design of the process system to which it is connected. Samplers are located as close as practical to the process stream such that sampic line interference will be insignificant.

5.5.2.1 Liquid Monitoring Systeme 5.5.2.1.1 Liquid Radioactive Waste Discharge Monitoring System (PRM-9)

PRM-9 monitors the activity of liquids discharged to the Plant liquid radioactive waste discharge header. The header receives liquid wastes from Plant systems that potentially contain radioactive J

nuclides and constitutes the path by which planned releases occur from the LRWS. Section 5.3 l

provides a description of the Liquid Radwaste System. l The monitoring system consists of a NaI gamma scintillation detector mounted externally on the ,

process piping. The radiation levels are displayed on a log ratemeter at the Plant liquid l radioactive waste discharge header. The ratenteter provides alert and high alarms which are l annunciated in a Control Room PC station. The high alarm setpoint is established in accordance l with the Trojan Nuclear Plant Offsite Dose Cr.lculation Manualm. Upon receipt of a high alarm or circuit failure signal, the Plant discharge header automatic isolation valve will close to immediately discontinue discharge flow to the discharge and dilution structure. The system response and isolation valve location are such that high activity ligtiids will not be released before the valve closes.

i 5.5.2.1.2 Liquid Sample Collection System f The Liquid Sample Collection System is provided to allow sampling at the intake structure and at the discharge and dilution structure for evaluation of the adequacy ofinplant monitoring.

O V 5.5-3 Revision 7 i

s

-)

The sampler at the discharge and dilution structure consists of a head tank which maintains constant level above the solenoid valve. The solenoid valve is provided with a timer to allow for ,

automatic sampling at adjustable frequencies and duration. The total sample volume collected can be manually adjusted by changing the time interval between resetting of the sampling timer and by adjusting the outlet throttle valve. I 5.5.2.2 Gas Monitoring Systems The following gas monitoring systems are provided:

~

(1) Containment Monitoring System (PRM-1)

(2) Auxiliary Building Vent Exhaust Monitoring System (PRM-2)

(3) SFP Vent Monitoring System (PRM-3)

Each of the gas monitoring systems consist of one or more detector channels dependent upon the service of the monitor. ,

The airborne particulate channel consists of a filter sampler and a beta scintillation detector that is capable of detecting the radioactivity emitted from the accumulated particulates on the filter.

The low-level and intermediate-level gas channels use a common shielded sample chamber containing a beta scintillation detector for low-level gaseous activity and a halogen-quenched Geiger-Mueller detector for intermediate-level activity. The two detector channels have one decade of overlap.

A three-way valve is provided at the sampler inlet to permit air purging of the sampler to facilitate background checks.

Revision 8 5.5-4 lI L _ __

The offline gas monitoring systems (PRM-1 and PRM-2) utilize sample pumps to draw the A.

Q necessary process gas samples through the sampler. The pumps provide a constant rate of sample -

flow irrespective of changes in flow resistance through the filter media.

Sample lines for offline samplers are fabricated of stainless steel. Sampling devices and procedures reflect the recommendations of ANSI 13.1-1%9, Guide to Sampling Airborne Radioactive Material in Nuclear Facilities *. An evaluation of sample line losses of radioiodine and air particulates was performed for PRM-1 and PRM-2. Line losses were found to be negligible. Each offline gas monitoring system is provided with a grab sample connection that can be used to obtain representative samples for laboratory analysis and with fixed-filter assemblies for collection of radioiodine, air particulates, and tritium.

The following gas monitoring method (s) are also provided:

(1) Radwaste Processing Building monitoring method p

(

The gas monitoring method utilized for the Radwaste Processing Building consists of an airborne particulate filter. A sample pump is used to draw the necessary gas sample through the sampler.

5.5.2.2.1 Containment Monitoring System (PRM-1)

PRM-1 monitors the air particulate activity levels in the purge exhaust duct when purging. The monitor quantitatively analyzes airborne activity released to the environs by the Containment Purge Exhaust System. Section 5.2.2 provides a description of the Containment Purge Exhaust System.

The Containment Monitoring System utilizes the air particulate channel (PRM-1 A).

A V 5.5-5 Revision 8

PRM-1 utilizes a single sample pump that draws a single gas stream through the airborne particulate monitoring moving filter. During containment purge operations, a sample is drawn l from and returned to the purge exhaust duct by an isokinetic sampling probe.

5.5.2.2.2 Auxiliary Building Vent Exhaust Monitoring System (PRM-2)

PRM-2 monitors the gaseous and air particulate activity levels released to the environment by the combined ventilation exhaust flows from the Fuel and Auxiliary Buildings. The system also monitors releases from the SFP ventilation exhaust and the vent collection header after they are diluted by the Fuel and Auxiliary Building ventilation exhaust flow. Section 5.2.3 provides a description of the Fuel and Auxiliary Building Exhaust System. ~

The Auxiliary Building Vent Monitoring System utilizes the following channels:

(1) Air particulate (PRM-2A)

(2) Low-range gas (PRM-2C)

(3) Intermediate-range gas (PRM-2D)

PRM-2 utilizes a single sample pump that draws a single gas stream in series through the airborne particulate monitoring moving filter and a low- and intermediate-level gas monitoring chamber.

Sample flow is drawn from and returned to the purge exhaust duct by an isokinetic sampling probe.

Revision 8 5.5-6

5.5.2.2.3 SFP Vent Monitoring System (PRM-3)

(T U -

PRM-3 monitors the gaseous activity levels released to the environment by the exhaust fans exhausting the SFP area of the Fuel Building. Diffusion of gaseous activity from the pool will generate airborne radioactivity in the area. Section 5.2.3 provides a description of the SFP area ventilation system.

PRM-3 consists of a shielded Geiger-Mueller detector mounted inside the common vent duct exhausting all Fuel Building spaces adjacent to the SFP.

5.5.2.2.4

~

Deleted l

5.5.2.2.5 Radwaste Processing Building Monitoring Method This method utilizes a single sample pump that draws 'a single gas stream through the airborne

( particulate monitoring filter. Sample flow is drawn from and returned to the exhaust duct by a

~

sampling probe. The sample is drawn downstream of the HEPA filter in the exhaust duct.

5.5.2.3 Annivtical Procedures Samples of process and effluent gases and liquids will be analyzed in the laboratory by the following techniques: ,

(1) Gross beta counting.

(2) Gross alpha counting.

(3) Gamma spectroscopy.

t 5.5-7 Revision 8 1

1

(4) Liquid scintillation counting.

(5) Radiochemical separations.

Gross beta analyses are performed with a thin-window proportional counter. Gross alpha analyses are performed with a solid-state alpha detector and scaler or thin-window proportional i counter. Sample volume, counting geometry and counting time are chosen to achieve the requisite measurement sensitivities.

Gamma spectrometry is used for isotopic analysis of samples. Complex gamma spectra are

~

resolved and analyzed by computer techniques.

Gaseous tritium concentrations are determined by adsorption on silica gel. Liquid samples for tritium analysis are purified prior to analysis by either passing the samples through mixed-bed ion-exchange columns or by distilling the samples, or both.

O Radiochemical procedures are available for the analysis of Sr-89'and Sr-90, 5.5.2.4 Calibration and Maintenance 5.5.2.4.1 Radiation Monitoring Systems The radiation monitors have been calibrated by the manufacturer. The manufacturer's calibration is traceable to certified NBS/NIST or commercial radionuclide standards. Following repairs or modifications, the monitors will be recalibrated at the Plant with the secondary radionuclide standards.

Revision 8 5.5-8 s

5.5.2.4.2 Laboratory' Radiation Detectors Counting efficiencies of all laboratory radiation detectors have been determined with certified radionuclide standards.

A periodic calibration check is performed to check the efficiency of "in use" laboratory radiation detectors. When the detector efficiency falls out of the three-sigma control limit, the detector performance will be evaluated and will be recalibrated in a timely manner. If the evaluation finds the detector performance to be unacceptable, the detector will be recalibrated prior to use. The detectors are recalibrated following repairs or modifications.

5.5.3 EFFLUENT MONITORING AND SAMPLING Enluent sampling of all principal radioactive liquid and gaseous effluent paths will be conducted on a regular basis to determine release rates to the environment and to verify the adequacy of effluent processing. The sample and analysis schedules and the offsite radiological monitoring program are described in the Trojan Nuclear Plant Offsite Dose Calculation Manual *. The schedules cc nply with the NRC positions glescribed in Regulatory Guide 1.21W.

! 5.5.4 PROCESS MONITORING AND SAMPLING Process sampling and monitoring is used to monitor activity levels within various Plant systems.

The sampling frequency, type of analysis, analytical sensitivity and the purpose of the sample are

~

summarized in Table 5.5-1 for each liquid process sample location.

b s 5.5-9 Revision 8

5.9 REFERENCES

REFERENCES FOR SECTION 5.1

1. Trojan Nuclear Plant Finni Safety Annivsis' Report, Amendment 19 (December 1992).

REFERENCES FOR SECTION 5.2 l

1. Trojan Nuclear Plant Finni Safety Analysis Report, Amendment 19 (December 1992).

2. PGE-1021. Trojan Nuclear Plant Offsite Dose Calculation Manual, Amendment 9 (February,1993).

i REFERENCES FOR SECTION 5.3

1. PGE Agreement With intervenors (May 1972).

2. Trojan Nuclear Plant Finni Safety Analysis Reoort, Amendment 19 (December 1992).

REFERENCES FOR SECTION 5.4

1. Trojan Nuclear Plant. Finni Safety Annivsis Reoort, Amendment 19 (December 1992).

~

REFERENCES FOR SECTION 5.5

1. Trojan Nuclear Plant Final Safety Analysis Report, Amendment 19 (December 1992).

2. PGE-1021. Trojan Nuclear Plant Offsite Dose Calculation Manual, Amendment 9 (February,1993).

3. Guide to Samnling Airborne Radioactive Material in Nuclear Facilities, ANSI 13.1-1969 l 5.9-1 Revision 8

l 4. Measurine. Evaluatine and Reoortine Radioactivity in Solid Wastes and Releases of Radioactive Materials in Lionid and Gaseous Effluents from Light-Water-Cooled Nuclear Plants, Regulatory Guide 1.21, Revision 1. U.S. Nuclear Regulatory Commission.

REFERENCES FOR SECTION 5.6 1.

Trojan Nuclear Plant Final Safety Analysis Report, Amendment 19 (December 1992).

REFERENCES FOR SECTION 5.7

1. Trojan Nuclear Plant Final Safety Analysis Report, Amendment 19 (December 1992).

l 2. Standards for Protection Against Radiation, Code of Federal Regulations, Title 10, Part 20.

l 3. Domestic Licensing of Production and Utilization Facilities, Code of Federal Regulations, Title 10, Part 50.

l 4. Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will be as Low as is Reasonably Achievable, Regulatory Guide 8.8, Revision 3. U. S. Nuclear Regulatory Commission (June 1978).

l 5. Operating Philosophy for Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable, Regulatory Guide 8.10, Revision IR. U. S. Nuclear Regulatory Commission.

REFERENCES FOR SECTION 5.8

1. Troian Nuclear Plant Final Safety Analvsis Report, Amendment 19 (December 1992).

Revision 8 5.9-2

~

! { ' I! l ll N

EOE GIR OC RTU A

SDRT I

D&S T

N LU C IU l

q T

N S E R E

V 8

i n

o s

cc IL

_ UT FLLI E i v

e N

OR1 F

E F R IO9 TT -

AIM <

INR DOP 1 }- }-

AIt - u R R #v LZ E

W g? L

' t AI NL OA 7 5h? '

- - IR TE

- - PN E

OI M

E D

m l t e

0 s y _

ES S TK e SN t AA s WT 1 S a DR F EO TT l

u P

W AI EN e RO T v TW W i t

W c T

r' a l A o O T S N R E E U

LLI T

i d

R a

M@N F I F d i

u q

i L

1

~_ -

S 3 E P T K M 5

~ SN U A TA P e W r T

Y N O u T I W g R A D i IR DD F

• N

/

N

/

o

! I {

6.2 FUEL HANDLING ACCIDENT Authorized fuel handling activities in the Fuel Building include moving spent fuel assemblies l within the SFP and loading spent fuel and other materials into Transfer and Storage Casks. l This sectio'n addresses the fuel handling accident associated with moving spent fuel assemblies l withm the SFP. Fuel handling accident scenarios associated with transferring spent fuel l between the SFP and Transfer and Storage Casks are provided in LCA 237 and LCA 246.

l A spent fuel assembly is a potentially significant source of radioactive releases due to the concentration of radionuclides inside the fuel rods and the potential release pathway.

6.2.1 ASSUMPTIONS OR CONDITIONS The possibility of a fuel handling accident is very remote due to the administrative controls and physical limitations imposed on fuel handling operations (See Sections 3.3.1 and 4.2).

Fuel transfer is conducted in accordance with prescribed procedures under direct surveillance of a Certified Fuel Handler trained in nuclear safety. Irradiated fuel is prohibited from being transferred into the reactor vessel or into containment.

Fuel handling manipulators and hoists are designed so that fuel remains below a position where the water depth provides adequate shielding for protection of personnel. The fuel handling manipulators, cranes, trollies, bridges and associated equipment above SFP are

- designed with protective features to preclude fuel damage. The facility is designed for transfer and handling of only one fuel assembly at a time, and movement of equipment when handling the fuel is restricted to low speeds.

These design safety features, in conjunction with the afoiementioned administrative controls, support the conclusion that the probability of a fuel handling accident is very small. Even if a fuel assembly was dropped or struck, the mechanical characteristics of the assembly indicate that not all of the fuel rods in the assembly would lose integrity. Nevertheless, the fuel 6.2-1 Revision 8

i handling accident analysis assu'mes that the cladding of all fuel rods in one assembly break and release all the gaseous fission products in the gas space of the fuel rods. The fraction of the fission products assumed to migrate from the fuel matrix to the gap and plenum regions and available for immediate release following clad damage is consistent with Regulatory Guide (RG) 1.25. Compared to the quantity immediately released, all subsequent activity releases were considered to be negligible.

J The analysis of activity released from the SFP resulting from a fuel handling accident takes a conservative approach to the evaluation of radiological consequences by assuming that the fuel assembly with the peak fission product inventory is the one damaged. This assembly is assumed to have had the highest peaking factor in the core and is assigned a conservative peaking factor of 1.65 per RG 1.25. Du: to the permanently defueled condition of the facility and decay time since last reactor operation the fission product inventory (especially of the short lived daughter products) is significantly less than that assumed for an operating plant.

Analysis of the activity released from the SFP water resulting from the postulated fuel handling accident has been based on the fission product source and release assumptions of RG 1.25.

The analysis assumes a conservative decay time of six months from the end of power operation l , (November 1992). Conservative values from Table 6.0-1 are used for the analysis. All of the gap activity in the damaged rods is assumed to be released to the SFP water and consists of 10 percent of the total noble gases other than Kr-85,30 percent of Kr-85 and I-129. Other iodine species have short half-lives and have decayed sufficiently so that they do not contribute to the dose consequences.

The calculational method used to develop the fuel handling accident analysis is consistent with the methods of Section 6.0.2. Code inputs were based on the inputs listed in Table 6.2-1.

l Radionuclide inventory was based on RG 1.25 gap fractions of core iodine and noble gas inventories from Table 6.0-1 corrected for the peak assembly. The accident assessment Revision 8 6.2-2

{

dilution factors were established values obtained from Table 6.0-5 for the site boundary, and from a calculation for the control room ventilation system intake (Reference 3). .

l The I-129 inventory present in the peak assembly at the time of Plant shutdown was determined to be 16.6 mci. This value was obtained by taking the total core inventory of 1.943 Ci at shutdown, dividing by the total number of assemblies in the core to obtain an average assembly value, then multiplying by the radial peaking factor of 1.65. The amount of -

I-129 available for release from the assembly was derived by multiplying this value by the 30 percent gap fraction from RG 1.25. Assuming a pool iodine decontamination factor of 100, the result was an airborne release of 50 pCi of I-129. This results in a maximum site boundary location thyroid dose of 0.6 mrem. -

6.2.2 Dose Results Results of the dose model for the fuel handling accident are shown with the 10 CFR 100 exposure limits in Table 6.2-2. From these results, it is concluded from the analysis that the  ;

whole body dose and inhalation thyroid dose are ., mall fractions of the 10 CFR 100 limits at the 662 meter exclusion area boundary and at the Control Building CB-2 ventilation system intake. The dose at the CB-2 intake is used'as a worst-case dose for personnel working at the facility; in addition it serves as the dose for control room habitability - the criteria for this is

. provided in 10 CFR 50 Appendix A General Design Criterion 19.

O '

6.2-3 Revision 8

6.4 REFERENCES

REFERENCES FOR SECTION 6.0

1. Calculation of Distance Factors for Power and Test Reactor Sites, TID-14844. l J. J. DiNunno, et al (March 1%2).

2. Summarv of Fission Product Yields for U-235. U-23.8. Pu-239. nnd Pu-241 at Thermal l Fission Socctrum and 14 Mev Neutron Energies, APED-5398. M. E. Meek and B. F. Rider (March 1%8).

3. Fission-Product Releace from UO2. D. F. Toner and J. S. Scott. Nuc. Safety 3, No. 2; l l p 15-20 (December 1%1).

4. Urnnium Dioxide Pronerties and Nuc1*nr Aoplientions. ,J. Belle. Naval Reactors, l Division of Reactor Development United States Atomic Energy Commission (1%1).

5. A Suonested Method for Calenistine the Diffusion of Rndinactive Rare Gas Fission l Prodticts From UO2 Fuel Elements, DCI-27. A. H. Booth (1957).

6. Reoort of ICRP Committee II Permicnible Dose for Internal Rndiation 1959. Health j, Physics, 3; p 30,146-153 (1960) .

O 7. Table of Isotopes, 6th ed. C. M. Iederer, et al (1%8).. .l j

8. Assumntions Used for Evaluntino the Potentint Radiological Conseauences of a Loss of ,

Coolant Accident for Pressurized Water Reactors, Regulatory Guide 1.4. Directorate of J Regulatory Standards, U.S. Atomic Energy Commission (June 1973).

9. Calculation of Distance Factors for Power and Test Reactor Sites, TID-14844. . l J. J. DiNunno, et al. U.S. Atomic Energy Commission (March 1%2).

10. EMFRAI D - A Program for the Calculation of Activity Relences and Potential Doses l from a Pressurized Water Reactor Plant. W. K. Brunot. Pacific Gas & Electric Company (October 1971).

11. PREL - A Program for the Calculation of Activity Release from a Reactor System. l WCAP-7461. W. K. Brunot, C. L. Beard and R. J. Lutz. Westinghouse Electric Corporation, Proprietary (April 1971).

12. WEDOSE - A Program for the Calculation of Potential Off-Site Doses from a Reactor l System, WCAP-7460. T. K. Shen, W. K. Brunot, and R. J. Lutz. Westinghouse Elect-ic Corporation, Proprietary (April 1971).

6.4-1 Revision 8

)

l

13. Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, Regulatory Guide 1.4. Directorate of Regulatory Standards, U.S. Atomic Energy Commission (June 1973) l 14 Meteorology and Atomic Energy, TID-24190. D. H. Slade. U.S. Atomic Energy Commission (1972).

REFERENCES FOR SECTION 6.1

1. Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure, Regulatory Guide 1.24 (Safety Guide 24). Directorate of Regulatory Standards, U.S. Atomic Energy Ccmmission (March 23,1972).

l 2. PGE Calculation RPC 93-025. Basic for Fuel Handling Accident as Limiting - Case -

Accident.

3. Seismic Design Classification, Regulatory Guide 1.29, Revision 3. U.S. Nuclear Regulatory Commission (September 1978).

REFERENCES FOR SECTION 6.2

1. Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel 0 Handling Accident in the Fuel Handling and Storage Facility for Boilin'g and Pressurized Water Reactors, Regulatory Guide 1.25. U.S. Atomic Energy Commission (March 23,1972). '

l 2. PGE Calculation RPC 93-025. Basis for Fuel Handling Accident as Limiting-Case Accident.

l 3. PGE Calculation RPC 93-003. Doses at Existing & Prooosed Site Boundarv from Fuel Handling Accident after 6 Months Decay, Revision 1 (March 9,1993).

i l 4. License Change Application (LCA) 237. Spent Fuel Cask Loading in the Fuel Buildmg. -

l Transmitted by PGE letter, VPN-005-99, dated January 7,1999.

I l 5. LCA 246. Soent Fuel Cask Loading in the Fuel Building - Contingency Fuel Unloading l to the Fuel Pool. Transmitted by PGE letter, VPN-006-99, dated January 27,1999.

Revision 8 6.4-2

---

_-----a

REFERENCES FOR SECTION 6.3

1. PGE Calculation TC-720. Troian Soent Fuel Pool Heatup and Boil Down Rates, l Revision 4 (November 12,1996).

2. PGE Calenistion RPC 93-24. Dose Rate Vs. Water Heioht Over Soent Fuel in Pool, l Revision 0 (September 16,1993).

O .

l O Revision 8 6.4-3

7.1.2.2 Supporting Organintiom The Supporting Orgamzations 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 Mananer. Knoineerina/Decommissionino - reports to the General Manager, Trojan Plant, and is responsible for engineering assistance, decommissioning, fire protection, and design control of Trojan. Under the direction of the General Manager, Engineering / Decommissioning, engineering support is provided for safe, efficient operation of Plant equipment, drawing and design document preparation, and plant modifications.

Manager. Decommicsioning Planning - responsible for the development of a decommissioning plan for the Trojan Nuclear Plant including the transition to dry fuel storage.

Manager. Engineering - responsible for engineering assistance, modifications, maintaining Plant configuration data and documentation, oversight and direction of the Fire Protection Program and Civil / Seismic Program. Also responsible for providing technical support for Plant Operations and Maintenance.

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

General Manager. Plant Support and Technical 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.

7.1-3 Revision 6

)

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

responsible for the administration of the Plant Security Organization, implementation of i 1

.the security program, and the Fitness-for-Duty Program. This responsibility includes

]

interac9on 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').

k l Manager. Nuclear Regulatory Affairs - responsible for all acti/ities required to maintain the permits and licenses required for the Plant including producing, maintaining, and interpreting licensing documents.

O Manager. Cost Control - responsible for cost control and procurement of equipment 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 Pivgram for each area that is audited, and for supporting line management in event analysis, 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 8 7.1-4 1

. I L. 4

- _-_______-_-______-____-__-________________-__U