ML20203B034
| ML20203B034 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 11/26/1997 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20203B032 | List: |
| References | |
| NUDOCS 9712120314 | |
| Download: ML20203B034 (103) | |
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O TABLE OF CONTENTS G-DEFUELED SAFETY ANALYSIS REPORT CHAPTER
1.0 INTRODUCTION
AND SUMMAR's Section Title . Page
1.0 INTRODUCTION
AND
SUMMARY
. . . . . . . . . . . . . . . . . . . . 1.0-1 1.1 Introduction . . ................................... 1.1-1 1.2 General Plant Description . ........................... 1.2-1 1.2.1 Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.2 Fuei llandling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-2 1.2.3 Radioactive Waste Treatment Systems . . . . . . . . . . . . . . . , , . . . 1.2-2 Identification of Agents and Contractors . . . . . . . . . . . . . . . . . . . 1,3 1 A 1.3 V
1.3.1 Design of the Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.4 Exclusions from and Exemptions to Certain Parts of Title 10 of the Code of Federal Regulations (10 CFR) , , . . . . . . . 1.4 1 1.4.1 Exclusions from Certain Parts of 10 CFR . . . . . . . . . . , . . . . . . . 1,4-1 1.4.1.1 10 CFR 26, Fitness for Duty Program . . . . . . . . ......... 1.4-1 1.4.1.2 10 CFR 50.44, Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors . . . . . . . . . . . . 1,4-1 1.4.1.3 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors . . . . 1.4-2 1.4.1.4 10 CFR 50.48, Fire Protection . . . . . . . . . . . . . . . . . . . . . . 1.4-2 1.4.1.5 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1,1979. . . 1.4-2 1.4.1.6 10 CFR 50,49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants . . . . . 1.4-2 1.4.1.7 10 CFR 50.55a, In-Service inspection Requirements . . . . . . . . . 1.4-3 1.4.1.8 10 CFR 50.60, Acceptance Criteria for Fracture Prevention Measures for Light-Water Nuclear Power Reactors . . . . . . . . . 1.4-3 1.4.1.9 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events . . . . . . . 1.4-3 i Revision 5 9712120314 971126 PDR ADOCK 05000344 W PDR y- m e..-n n <-m - +9-, - +, ,,, ,,, . , _ _ , _ _ . _ _ , _ _ , _ _ _ , , _ _ , _ , , , _ _ , _ _ _ _ _ _ _ _ _ _ _ , . _ _ _ _ , _ _ _ _ , , , .
CHAPTER 1.0 g INTRODUCTION AND
SUMMARY
W Section ,
Title Page 1.4.1.10 10 CFR 50, Appendix G, Fracture Toughness Requirements . . . . 1.4-3 1.4.1.11 10 CFR 50, Appandix II, 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(c), Maintenance of Record, Making of Reports . . . 1.4-5 1.4.1.15 10 CFR 70.24, Criticality Accident Requirements . . . . . . . . . . . 1,4-5 1.4.2 Exemptions to 10 CFR Related to the Permanently Defueled Condition . . . . . . . . . . . . . . . . . . . . . . . 1.4-5 1.4.2.1 10 CFR 50.54(o) and Appendix J, Primary Reactor Containment Irakage Testing for Water-Cooled Power Reactors . . . . . . ......................... 1,4-5 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, g
" Emergency Plans," ............................ 1.4-6 W l.5 Material Incorporated by Reference . . . . . , , . . . . . . . . . . . . . . . 1.5-1 O
- Revision 5 ii
d r CHAPTER 2.0
('-} SITE CHARACTERISTICS '
Section Title Page
- 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 S ite Area M ap . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 . . . . . . . 2.14 2.1.2.3 Arrangements for Traffic Control . . . . . . . . . . . . . . . . . . . . . 2.1-5 2.1.3 Population Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-6 2.1.3.1 Population Within 19 Miles . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-6 O 2.1.3.2 Population Between 10 and 50 Miles . . . . . . . . . . . . . . . . . . . 2.1-8 V
2.1.3.3 Transient Population . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-8 2.1.3.4 Low Population Zone . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-9 2.1.3.5 Population Center . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.1-10 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 Mate: rials . . . . . . . . . . . . . . . . . , 2.2-4 2.2.2.2 Pipelines ..................................... 2.2-5 2.2.2.3 Waterway s . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . 2.2-6 2.2.2.4 A i rpo rts . . . . . . . . . . . . . . . . . . . . . . . . . . . , , , . . . . . . . . . 2.2-6 2.2.3 Evaluation of Potential Accidents ....................... 2.2-7 2.2.3.1 Ex plos io ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 V 2.2.3.5 Oil or Corrosive Liquid Spills in River . . . . . . . . . . . . . . . . . . 2.2-21 iii Revision 5 4
I CllAPTER 2.0 SITE CilARACTERISTICS g
Section Title .fagc.
2.2.3.6 Cooling Tower Collapse . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.2 22 2.3 hicteomlegy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3 1 2.3.1 Regional Climatology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3 1 2.3.1.1 Gene ral Climate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3 1 2.3.1.2 Regional hieteorological Conditions for Design and Operation 11ases . . . . . . . . . . . . . . . . . . . . . . . . 2.3 1 2.3.2 Local hieteorology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.3 2 2.3.2.1 Noimal and Extreme Values of hieteorological Parameters . . . . 2.3 2 2.3.2.2 Potentialinfluence of the Plant and its Facilities On Local hieteorology . . . . . . . . . . . . . . . . 2.3-4 2.3.2.3 Local meteorological Conditions for Design and Operation llases . . . . . . . . . . . . . . . . . . . . . . . . 2.35 2.3.3 Onsite hieteorological hicasurements Program . . . . . . . . . . . . . . . 2.3-6 g 2.3.3.1 Past hieteorological Facility Operations . . . . . . . . . . . . . ... 2.3 6 W 2.3.3.2 hicasurements . . . . . . ....................... ... 2.3-7 2.3.4 Diffusion Estimates ................................ 2.3 8 2.4 Ilydrologic_lingineering ............................. 2.4-1 2.4.1 Ilydrologic Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 2 2.4.1.1 Site and Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-2 2.4.1.2 Ilydrosphere . . . . . . . . . . . . . . .................... 2.4-2 2.4.2 Floods ....................................... 2.4-3 2.4.2.1 Fl o<x! l l ist o ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4-3 2,4.2.2 Flood Design Considerations , , , , . . . . . . . . . . . . . . . . . . . 2,4-4 2.4.2.3 Effects of local Intense Precipitation . . . . . . . . . . . , . . . . . . . 2.4-5 2.4.3 Probable hiaximum Flom.1 cf Streams and Rivers . . . . . . . . . . . . . 2.4-5 2.4.3.1 Probable hiaximum Precipitation . . . . . . . . . . . . . . . . . . . . . . 2,4 5 2.4.3.2 Precipitation Losses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 8 2.4 3.3 Runoff hiodel . . . . . . . . . . . . . . ................... 2.4 9 2.4.3.4 Probable hiaximum Flood Flow . . . . . . . . . . . . . . . . . . . . . . 2.4 12 Revision 5 iv
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CilAPTER 2.0 SITE CilARACTERISTICS Section Title lagc.
2.4.3.5 Water Ixvel Determinations . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 18 2.4.3.6 Coincident Wind Wave Activity . . . . . . . . . . . . . . . . . . . . . . 2.4 19 2.4.4 Potential Dam Failures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 20 2.4.4.1 Seismically Induced Dam Failure . . . . . . . . . . . . ........ 2.4 20 2.4.4.2 Volcanically Induced Dam Failure . . . . . . . . . . . . . . . . . . . . . 2.4 22 2.4.4.3 Spirit Lake Illockage Failure ........................ 2.4-25 2.4.5 Probable Maximum Surge Flooding . . . . . . . . . . . . . . . . . . . . . . 2.4 29 2.4.5.1 Surge W ater levels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2,4 29 a.4.5.2 Resonance ........................... ........ 2.4 30 2.4.6 Pr6able Maximum Tsunami Flooding . . . . . . . . . . . . . . . . . . . . 2,4 31 2.4.7 Ice Effects ........... .......................... 2.4 31
/~~'N 2.4.8 Cooling Water Canals and Reservoirs . . . . . . . . . . . . . . . . . . . . . 2.4 32 e <
U 2.4.9 Channel Diversions ................................ 2.4 32 2.4.10 Flooding Protection Requirements . . . . . . . . . . . . . . ........ 2.4 32 2.4.11 Low Water Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 33 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of I lyid Effluents in Surface Waters . . . . . . , . . . . . 2.4-33 2.4.13 G rou nd wa t e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 34 2.4.13.1 Description and Onsite Use . . . . . . . , , . . . . . . . . . . . . . . . . 2,4-34 2.4.13.2 Sou rc e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.4 35 2.4.13.3 Accident F.ffects . . . . . . . . . . . . . ................... 2.4-36 2.5 Geology. Seismology and Geotechnical Engineering . . . . . . . . . . . 2.5-1 2.5.1 11asic Geologic and Seismic Information . . . . . . . . . . . . . , . . . 2.5 1 2.5.1.1 Regional Geology . . . . . . . . . . . . . . . . . . ............. 2.5 2 2.5.1.2 S ite Geology . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -2.5-4 s.
N v Revision 5
i CllAPTER 2.0 G4 SITF CilARACTERISTICS Seclinn Title Page 2.5.2 Vibra:ory Ground Motion . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 8 2.5.2.1 Se i s m ic i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 8 2.5.2.2 Geologic Structures and Tectonic Activity . . . . . . . . . . . . . . . . 2.5 8 2.5.2.3 Maximum Eanhquake Potential . . . . . . . . . . . . . . . . . . . . . . . 2.5 9 2.5.2.4 Seismic Margin Earthquake, L.fe Shutdown Earthquake and Operating Basis Earthquake . . . . . . . . . . . . . . . . . . . . . 2.5 11 2.5.3 Su r face Faulting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 13 2.5.3.1 Geologic Condition of the Site . . . . . . . . . . . . . . . . . . . . . . . 2.5-13 2.5.3.2 Investigation for Capable Faults . . . . . . . . . . . . . . . . . . . . . . 2.5 14 2.5.3.3 Description of Faults . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-16 2.5.4 Stability of Subsurface Materials and Foundations . . . . . . . . . . . . . 2.5 28 2.5.4.1 Foundation Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 29 2.5.5 Stability of Slopes ................................. 2.5 33 2.5.6 Volc a nol og y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-33 2.5.6.1 Volcanoes in General Area . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-34 2,5.6.2 Possible llazards of the Cascade Volcanoes . . . . . . . . . . . . . . . 2.5 38 2.5.6.3 Ilazards from Future Volcanic Activities . . . . . . . . . . . . . . . . . 2.5 39 2.5.6.4 Sumno .y and Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5-47 2.6 Refuentes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.6-1 Revision 5 vi
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V CilAPTER 3.0 1%CILITY DESIGN Section Title .hge.
3.0 FACILITY DESIGN ............................... 3.01 3.1 Su mma rv . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11 3.1.1 Conformance with NRC General Design Criteria . . . . . . . . . . . . . 3.1-1 3.1.2 Clastification 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 3.1.4 Water level (Flood) Design . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 13 3.1.5 Missile Protection Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-14 3.1.5.1 Mlasile Selection and Description . . . . . . . . . . . . . . . . . . . . . 3.1-14 O
V 3.1.6 Seismic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 15 3.1.6.1 Seismic Design Parameters . . . . , . . . . . . . . . . . . . . . . . . . . 3.1 15 3.1.6.2 Seismic Syste.n Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-19 ,
3.2 Soent Fuet Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2 1 3.2.1 Control. Auxiliary, and Fuel Building Complex . . . . . . . . . . . . . . 3.2-1 3.2.1.1 Desc ription . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... 3.21 3.2.1.2 Desig n Ba ses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-6 3.2.1.3 Applicable Codes, Standards, and Speci0 cations . . ... ... 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 Tecimiques . . . . . . . . . . . . . . . .................. 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 Des ig n Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-35 O) t v
vil Revision 5
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l CilAPTER 3.0 g FACILITY DESIGN W Section Title .hge.
3.2.2.2 Sy stem Des ig n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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< valuation for Beyond Design Basis Seismic Motions . . . . . . . . . . . . . . . . . . . . . . . 3.2-40 3.3 Au x iliarv Sy stems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3 3.1 Fuel liandling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 1 3.3.1.1 De s ig n lla se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 1 3.3.1.2 Sy stem Desc ription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-2 3.3.1.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-4 3.3.2 SFP Cooling and Demineralizer System . . . . . . . . . . . . . . . . . . . 3.3-5 3.3.2.1 De sig n B a se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 5 3.3.2.2 System Desc ription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-6 3.3.2.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 8 3.3.3 Component Cooling Water System . . . . . . . . . . . . . . . . . . . . . . . 3.3 8 3.3.3.1 De sig n Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-9 3.3.3.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 9 3.3.3.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 11 3.3.4 Service Water System . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . 3.3-12 3.3.4.1 De sig n B ases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-12 3.3.4.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 13 3.3.4.3 Desig n Evaluation . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . 3.3 14 3.3.5 Comprested Air System ............................. 3.3 15 3.3.6 Boric Acid Batch Tank . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 15 3.3.7 Makeup Water Treatment System . . . . . . . . . . . . . . . . . . . . . . . 3.3-16 3.3.8 Equipment and Floor Drain Systems . . . . . . . . . . . . . . ...... 3.3 16 Revision 5 viii
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CIIAPTER 3.0 FACILITY DESIGN :
1 Section ._
Title .hge. l 1
3.3.9 Plant Discharge and Dilution Stmeture . . . . . . . . . . . . . . . . . . . . 3.3 17 3.3.10 Primary Sampling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 18 3.3.11 Fire Protection System and Program . . . . ................ 3.3 18 3.3,12 Control Room IIabitability . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3 19 3.3.13 Seismic Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-19 3.4 Electric Powtr . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.1 Of fsite Power System . . . . . . . . . . . . . . . . . . ............ 3.4-1 3.4.1.1 De sc ri pt ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.1.2 A na ly s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4 2 3.4.2 Onsite Power Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4 2 pd 3.4.2.1 Desc ription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-2 3.4.2.2 A na l y s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . 3.4-4 3.5 Compliance with NRC Regulatorv Guides . . . . . . . . . . . . . . . . . . 3.5 1 3.6 Refe re nsc2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-1 l
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CllAPTER 4.0 ,
OPERATIONS i Section Title Page 4.0 O P E RATION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.0-1 4.1 Operadon Desc ription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 . . . . . . . . . . . . . . . . . . . . . . 4.1-4 4.2 S ne nt Fu el l l and li ng . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . 4.2 1 4.2.1 Spent Fuel Receipt, llandling, and Transfer . . . . . . . . . . . . . . . . 4.2-1 4.2.1.1 4.2.1.2 Fonctional Description . . . . . . . . . . . . . . . . . . . , . . . . . . .
Sa fe ty Featu res . . . , . . . . . . . . , . . . . . . . . . . . . . . . . . . .
4.2-1 4.2 2 h
4.2.2 Spent Fuel Storage . . . . . , , . . . . . . . . . . . . . . . . . . . . . . . . 4.2-3 4.3 Spent Fuel Cooling and Support Systenu . . . . . . . . . . . . . . . . . 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 liigh Spent Fuel Pool Level ........................ 4.3-4 4.3.1.5 Safety Criteria and Assurance . . . . . . . . . . . . . . . . . . . . . . . 4.3-4 4.3.2 Electrical Distribution . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . 4.3-5 4.3.3 S spport Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.3-5 4.4 Control Room Area . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.4 1 4.5 Re fe re nce s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.5 1 x
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Cl{ APTER 5.0 RADIATION PROTECTION Title .Eagc.
Section RADI ATION PROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . 5.0 1 5.0 i
Sourcc Terms ........................ ..... .. 5.I*1 5.1 o
Offgan Treatment ami Ventilation . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2 ,
Vent Collection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-1 5.2.1 De s i g n B a se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2 1 ,
5.2.1.1 System Description . . . . . . . . . . . . .....,........... 5.2-2
, 5.2.1.2 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-2 5.2.1.3 1
Containment Ventilation System . . . . . . . . . . . . . . . . . . . . . . . . 5.2-3 5.2.2 De s i gn B a se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-3 5.2.2.1 System Ikscription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-3 5.2.2.2 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-5 5.2.2.3 Fuel and Auxiliary Building Ventilation System . . . . . . . . . . . . . . 5.2 5 5.2.3 Des i gn Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2 5 5.2.3.1 5.2.3.2 System Description . . . . . . . . . . , , . . . . . . . . . . . . . . . . . . 5.2-5 ,
Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-6 5.2.3.3 5.2.4 Radwaste Processing Building Ventilation System . . . . . . . . . . . . 5.2-7 5.2.4.1 De s i g n B a se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-7 5.2.4.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2-7 5.2.4.3 Design Evaluation . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 5.2 8 5.3 Liquid Waste Treatment ami Retention . . . . . . . . . . . . . . . . . . . 5.3 1 5.3.1 Clean Radioactive Waste System . . . . . . . . . . . . . . . . . . . . . . . 5.3 1 ,
5.3.1.1 Design Bases ................................. 5.3 1 5.3.1.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-2 5.3.1.3 Design Evaluation . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . 5.3-4 (
5.3.2 Dirty Radioactive Waste System ....................... 5.3-5 5.3.2.1 Design Bases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3 5 5.3.2.2 System Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3 '
5.3.2.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.3-8 xi Revision 5
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5.4 Solid Was t es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-1 5.4.1 Desig n Base s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 Filte r liandling . . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . 5.4 3 5.4.2.3 Solid Waste s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-3 5.4.3 Design Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.4-4 5$ Process and Effluent Monitoring Systems . . . . . . . . . . . . . . . . . . 5.5 1 5.5.1 De sig n Base s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5 1 5.5.2 Syste m Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-2 5.5.2.1 Liquid hionitoring Systems . . . . . ................... 5.5-3 5.5.2.2 Gas hionitoring Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5-4 5.5.2.3 Analytical Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.5 8 5.5.2.4 Calibration and hiaintenance . . . . . . . . . . . . . . . . . . . . . . . . 5.5-9 5.5.3 Efauent hionitoring and Sampling . . . . . . . . . . . . . . . . . . . . . . 5.5 9 5.5.4 Process hionitoring and Sampling ...................... 5.5-10 5.6 Radiation Protection Program . . . . . . . . . . . . . . . . . . . . . . . . . 5.61 5.6.1 Radiation Protection Design Features . . . . . . . . . . . . . . . . . . . . 5.6-1 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 hionitoring 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 xii 9
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CilAPTER 5.0 RADIATION PROTECTION Section Title _Pagc_
5.6.3.3 Radioactive Materials Safety . . . . . . . . . . . . . . . . . . . . . . . . 5.69 5.7 Ensurine that Occupational Radiation Exposures ate a low as is Reasonably Achievable (ALARA) ..................... 5.7-1 5.7.1 Policy Considerations .............................. 5.7 1 5.7.2 Design Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7-1 5.7.3 Operational Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.7-1 5.8 Collective Dose Assessment .......................... 5.8 1 5.9 References .............................,....... 5.9 1 rx U
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l 6.0 Accident Anah313 ................................. 6.0 1 6.0.1 Fission Product Inventories . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0 1 6.0.1.1 Activities in the Core . . . . . . . . . . . . . . . . . . . ......... 6 0-2 6.0.1.2 Activities in the Fuel Pellet Cladding Gap . . . . . . . . . . . . . . . . 6.0-2 6.0.2 Radiological Evaluation Model . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-4 !
6.0.2.1 A s s u m pt io ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0-5 l 6.0 5 6.0.2.2 Whole ikx!y Dose ...............................
6.0.2.3 Thyroid Inhalation Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.0 7 6.0.2.4 Computer Code . ............................... 6.0 8 6.1 Radioactive ReleascJrom a Subsystem or Component .......... 6.1-1 j 6.1.1 Radioactive Gas Waste System leak or Failure . . . . . . . . . . . . . . 6.1-1 ;
6.1.1.2 Assumptions or Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1 2 l 6.1.1.3 Dose Re sul ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1 2 6.1.2 Postulated Radioactive Releases Due to Liquid Tank Failures . . . . . . . . .................... 6.1-3 6.1.2.1 Identincation of Causes and Accident Description . . . . . . . . . . . 6.1-3 6.1.2.2 Dose Re su l t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1-4 6.2 Fu e l l l a nd l i n g Acddent . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.2-1 6.2.1 Assumptions or Conditions ........................... 6.2 1 6.2.2 Dose Resul ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.2-3 6.3 Snent Fuel Pool Aggidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3 1 6.3.1 less of Spent Fuel Decay lleat Removal Capability . . . . . . . . . . . 6.3-1 6.3.1.1 Potential Events Resulting in less of Spent Fuel Decay lleat Removal Capability , , . . . . . . . . . . . . . . . 6.3 1 6.3.2 Loss of Forced Spent Fuel Cooling Without Concurrent SFP Inventory Loss . . . . . . . . . .............. 6.3-4 O
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t CIIAPTER 6.0 ACCIDENT ANALYSIS Section Title . fagc_
6.3.3 Loss of Forced Spent Fuel Cooling with Concurrent S FP inventory Loss . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.3 5 6.4 Re fe re nc e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.4-1 a
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CIIAPTER 7.0 CONDUCT OF OPERATIONS Section Title Page 7.0 CONDUCT OF OPERATIONS . . . . . . . . . . . . . . . . . . . . . . . . 7.0-1 7.1 PGE Organizational Stmeture . . . . . . . . . . . . . . . . . . . . . . . . . . 7.1 1 7.1.1 Management and Technical Support Organization . . . . . . . . . . . . . 7.1-1 7.1.2 Nuclear Division . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.1-1 7.1.2.1 Plant Organizations ............................. 7.1-1 1 7.1.2.2 Supporting Organizations . . . . . . . . . . . . . . . . . . . . . . . . . . 7.1-3 i 7.1.2.3 Review and Audit Organizations . . . . . . . .............. 7.1-5 l 7.2 Pl a nt Procellu re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2 1 7.2.1 Procedures Related to Nuclear Safety . . . . . . . . . . . . . . . . . . . . . 7.2 1 7.2 2 I 7.2.2 Nuclear Division Manual . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7.2.3 7.2.3.1 Pla nt Proced u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Administrative Orders . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
. 7.2-3 7.2-4
'h 7.2.3.2 Operating Instmetions . . . . . . ...................... 7.2-4 7.2.3.3 Pe riod ic Tes ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-4 7.2.3.4 Fuel llandling Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-4 7.2.3.5 Maintenance Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-5 7.2.3.6 Radiation Protection Procedures . . . . . . . . . . . . . .... ... 7.2-5 7.2.3.7 Chemistry Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-5 7.2.3.8 Plant Safety Procedures . . . . ..... ................. 7.2-6 7.2.3.9 Temporary Procedu r es . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.2-6 7.3 Iraining ....................................... 7.3 1 7.3.1 Training for Certified Fuel llandlers ..................... 7.3 1 7.3.2 Training for Plant Staff . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.3-1 7.3.2.1 General Employee Training . . . . . . . . . . . . . . . . . . . . . . . . . 7.3 .
7.3.2.2 Fire I3rigade Training . . . . . . . . . . . . . , . . . . . . . . . . . . . . . 7.3-2 7.3.2.3 Other Training Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.3 2 7.4 Emery crcy Pl an . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.4-1 Revision 5 xvi O
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CilAPTER 7.0 !
CONDUCT OF OPERATIONS .
I Section Title .Eagt.
7.5 Decommlutoning Plan .............................. 7.5 1 l 7.6 Troian Nucigar Plant Security Plan /Troian Nuclear Plant Security Force Training and Oualification Plan .................... 7.61-T 4
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CilAI'l'ER 8.0 TECllNICAL SPECIFICATIONS g D Title Page 8.0 TECilNICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . . 8.0-1 1
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CliAPTER 9.0 QUALITY ASSURANCE Title _Eagc.
SccliOD i
9.0 QUALITY ASSURANCE , . . . . . . . . . . . . . . . . . . . . . . . . . . 9.0 1 )
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O xix Revision S
LIST OF TABLES g DEFUEl ED SAFETY ANALYSIS REPORT Number Title 2.3 1 Trojan Cumulative Frequency Distribution of X /Q Values (September 1,1971 - August 31,1972) ;
2.3 2 Annual Average x /Q Values for Continuous Ground Level Releases (Trojan Site Data September 1,1972 - August 31,1974) 2.3 3 Annual Average Deposition Values for Continuous Ground-Level Releases (Trojan Site Data September 1,1972 - August 31,1974) 2.34 Annual Average x /Q, Deposition and Plume Depletion Factor at Site Boundaries and Offsite Exposure Locations for Ground Level Releases (Trojan Site Data September 1,1972 - August 31,1974) 2.3-5 hfaximum Annutar Sector, Terrain Adjustment Factors Derived From NUSPUF With Building Wake Adjustment Divided by NUSOUT for Standard Population Distances of 0.5 hiile to 4.5 hiiles 2.3-6 Terrain Adjustment Factors Derived fonn NUSPUF With Building Wake Adjustment Divided by NUSOUT for Special Instances 3.1-1 Design Wind leads 3.1-2 Capability of Structures to Withstand a Tornado 3.1-3 Locations of Gas Storage Tanks 3.1-4 Analyzed SFP Load Drops and hlissiles 3.1-5 Damping Values, Percent Critical Damping 3.1-6 Frequencies and biodal Effectiv '. Weights for Control, Auxiliary, and Fuel Building Complex 3.1-7 Comparison of hiaximum .%celerations from Time-liistory and Response Spectrum Analyses of the Unmodified Control Building Complex 3.2-1 1.ive Loads 3.2-2 Calculated Results - Fuel Building 3.2-3 Calculated Results - Auxiliary Building 3.2-4 Calculated Results - Control Building 3.2-5 Signi'icant Elevations in the Spent Fuel Pool 3.5 1 List of Pertinent Regulatory Guides for Defueled Status 4.1-1 Seismic hionitoring Instrumentation 4.3-1 Spent Fuel Pool Chemistry Specification and Sampling Schedule O
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DEFUFI Fn SAFETY ANALYSIS REPORT l
Numher Title 5.5 1 Radiological Analysis Summary of Liquid Process Samples ;
i 6.0 1 Core and Gap Activities Based on Full Power Operation for 650 Days Full i Power: 3565 MWt 6.0-2 Core Temperature Distribution ,
6,0-3 Breathing Rates 6.0-4 Physical Data for Isotopes 6,05 Accident Atmosplwric Dilution Factors
. 6.2 1 - Input Data for Calculation of Site Boundary Doses of a Fuel llandling Accident '
6.2 2 Resultant Doses from Fuel llandling Accident and Comparison with 10 CFR 100-6,3 1 Spent Fuel Pool Performance During loss of Forced Cooling 6.3 2 Dose Rates at Spent Fuel Pool at Reduced Water Levels ,
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LIST OF FIGURES g DEFUELED SAFETY ANALYSIS REPORI Number- Title 1.1-1 Facility Location 3.1 1 Design Response Spectra Operating Basis Earthquake 3.1-2 Design Response Spectra Safe Shutdown Earthquake 3.1-3 Synthetic Acceleration Time Ilistory Noimalized to 1.0 g 3.1-4 Comparison of the Acceleration Response Spectrum of the Synthetic Time llistory With the Trojan Design Spectnim for 1% Damping 3.1-5 Comparison of the Acceleration Response Spectrum of the Synthetic Time Ilistory With the Trojan Design Spectnim for 2% Damping 3.1-6 Comparison of the Acceleration Response Spectrum of the Synthetic Time llistory With the Trojan Design Spectnim for 5% Damping 3.1-7 Comparison of the Acceleration Response Spectrum of the Synthetic Time Ilistory With the Trojan Design Spectnim for 7% Damping 3.1-8 Finite Element Model - Control, Auxiliary, and Fuel Building 3.1-9 North - South llorizontal Response Spectra for Control Building Elevations 61 feet 0 inch and 65 feet 0 inch 3.1-10 North - South liorizontal Response Spectra for Auxiliary Building Elevation 61 feet 0 inch 3.1-11 North - South llorizontal Response Spectra for Fuel Building Elevation 61 feet 0 inch 3.1-12 North - South llorizontal Response Spectra for Control Building Elevation 117 feet 0 inch 3.1-13 North - South llorizontal Response Spectra for Auxiliary Building Elevation 117 feet 0 inch 3.1-14 East - West llorizontal Response Spectra for Control Building Elevations 61 feet 0 inch and 65 feet 0 inch 3.1-15 East - West llorizontal Response Spectra for Auxiliary Building Elevation 61 feet 0 inch 3.1-16 East - West llorizontal Response Spectra for Fuel Building Elevations 61 feet 0 inch 3.1 17 East - West llorizontal Response Spectra for Control Building Elevation 117 feet 0 inch 3.1-18 East - West Ilorizontal Response Spectra for Auxiliary Building Elevation 117 feet 0 inch 3.1-19 3 D Mod:1 for the Fuel Building Steel Superstructure 3.2-1 Fuel Building - Plan Elevation 45' O
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LIST OF FIGURES DEFUEL Ff) SAFETY ANALYRIM REPORT
! i Numher ' Thle
[ 3.2 2 Spent Fuel Pool- Typical Details 3.2 Auxiliary Building - Plan Elevation 45' !
t 3.2 4' Auxiliary Building Plan Elevation 45', Containment Abutment 3.2-5 Fuel Building - Plan Elevation 66' i i 3.2 6 - Auxiliary Building - Plan Elevation 61' Typical Section Through Auxiliary and Fuel Buildings :
3.2-7 3.2-8 Typical Steel Framing ;
3.2 9 Typical Steel Column Details l 1 3.2 10 Floor Plans Showing Modifications .
3.2 11 Control Building Floor Plan EL 45'-0" Showing Existing and Shear Walls
, 3.2 12 Control Building Floor Plan EL 61'-0" & 65'-0" Showing Existing and
! Shear Walls l 3.2 13 Control Building Floor Plan EL 77'-0" Showing Existing and Shear Walls 3.2-14 Control Building Floor Plan EL 93'-0" Showing Existing and Shear Walls 3.2-15 Equipment Location, P.cactor and Auxiliary Buildings Plan Below j(-
Ground Floor
.3.2 16 Equipment Location, Reactor and Auxiliary Buildings - Plaa Operating .
2 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 l Floor and Above 3.2 20 Equipment Location, R.: actor and Auxiliary Buildings - Section A A 3.2 21 Equipment Location, Reactor and Auxiliary Buildings - Sections B B, 4
D D, E E, F F _
, 3.2-22 Equipmem Location, Reactor and Auxiliary Buildings - Sections C-C and F F 3.2 23 Containment Structure Typical Details 3.2 24 Containment Stmeture Typical Liner Plate Details
, .t.2 25 Containment Structure Base Slab Bottom Reinforcing j 3.2 26 Containment Structure Base Slab Top Reinforcing 3.2 27 Containment Structure Wall Reinforcing 3.2 28 _ Containment Structun Dome - _
t
. 13.2-29 Containment Structure Typical Penetration Details ,
^
3.2-30 Containment Structure Prestressing Tendons at Equipment Hatch 3.3 1 Fuel Pool Cooling and Demineralizer System l xxiii - Revision 5
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I LIST OF FIGURES h DEFUELED SAFETY ANALYSIS REPORT l
Number Title l 3.3-2 Component Cooling Water 3.3-3 Service Water System ,
l l
4.2-1 Spent Fuel Storage Pool 4.2 2 Plant Arrangement Diagram of Fuel Cask hiovement Envelope 5.2 1 Gaseous Radioactive Waste System 5.2 2 Containment Purge Supply System (CS-1) 5.2 3 Containment Purge Exhaust System (CS-2) 5.2-4 Fuel / Auxiliary Building Ventilation Supply Syste. (AB-2) 5.2 5 Fuel / Auxiliary Building Ventilation Exhaust Syste .. (AB 3) 5.2-6 SFP Ventilation Exhaust System (AB-4) 5.2 7 Condensate Demineralizer Building Ventilation Exhaust System 5.3 1 Clean Radioactive Waste System 5.3.2 5.3 3 Dirty Radioactive Waste System Liquid Radwaste Demineralizers g
5.4-1 Solid Radioactive Waste System 6.1-1 Waste Gas Storage Tank Rupture Whole Body (Beta plus Gamma) Dose 6.3-1 Decay lleat Generated from Stored Fuel 6.3 2 SFP Ileatup 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 hiakeup versus Time After Reactor Shutdown 6.3 5 hiakeup Rate to hiaintain 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 Revision 5 xxiv
LIST OF EFFECTIVE PAGES DEFUFI FD SAFETY ANALYSIS REPORT Section Effective Paces Revision ,
. Title Page N/A key.O Table of Contents ! - xxiv Rev.5 ;
list of Effective Pages xxv - xix Rev.5 1.0 . 1.0 1 Rev.0 1.1 1.11 Rev.0 1.2 1.2 1 Rev.0 1.2 1.2 2 Rev.4 l.3 1.3-1 Rev.0 1.4 1.41 through 1,4-6 Rev.4 1.5 1.5-1 Rev. 5 Figure 1.1-1 N/A Rev.0 2.0 7.01 Rev.0 4 2.I 2. i-1 and 2.12 Rev 0 2.1 2 1-3 Rev.3 2.1 2.1-4 through 2.1-12 Rev.0 1 2.2 2.2-1 through 2.2 7 Rev.0 2.2 2.2 8 Rev.4 2.2 2.2 9 Rev.0 2.2 2.2-10 Rev.4 2.2 2.211 through 2.2-16 Rev 0 2.2 2.2 17 Rev.3 2.2 2.218 through 2.2-22 Rev.0 2.3 2.31 through 2.312 Rev.0 2.4 2.4-1 Rev.3 2.4 2,4 2 and 2.4-3 Rev.0 2.4 2.4-4 Rev.3 2.4 2,4 5 through 2,4-31 Rev.0 2.4 2.4 32 Rev.4 2.4 2.4 33 through 2.4-36 Rev, 0 2.5 2.51 through 2.5-12 Rev.0 2.5 2.5-13 Rev.4 2.5 2.514 through 2.5-47 Rev.0 2.6 2.6-1 through 2.6-10 Rev.O Tables 2.3-1 througn 2.3-6 N/A Rev.0 xxy Revision 5
1 LIST OF EFFECTIVE FAGES del!UELED SAFETY ANALYSIS REPORI I Section Effective Pages Ec.rision Figure 2.31 N/A Rev.O Figures 2.41 and 2.4 2 N/A Rev.0 l
3.0 3.0 1 Rev.0 3.1 3.1 1 through 3.1-5 Rev.0 3.1 3.16 and 3.1-7 Rev.5 3.1 3.18 through 3.1-28 Rev.0 3.2 3.2-1 Rev.0 3.2 3.2 2 Rev.4 3.2 3.2 3 through 3.2-7 Rev.5 3.2 3.2 8 through 3.2 34 Rev.4 3.2 3.2 35 through 3.2-41 Rev.5 3.3 3.3 1 Rev.5 3.3 3.3 2 through 3.3 3 Rev.0 3.3 3.3-4 Rev.4 3.3 3.3-5 through 3.3-7 Rev.0 3.3 3.3 3.3-8 3.3 9 Rev.4 Rev.3 h
3.3 3.310 and 3.311 Rev.5 3.3 3.3 12 Rev.4 3.3 3.313 through 3.3-19 Rev 5 3.4 3.4-1 Rev.0 3.4 3.4-2 through 3.4-4 Rev.4 3.4 3.4 5 through 3.4-6 Rev.0 3.5 3.5 1 Rev.0 3.6 3.6-1 Rev.0 3.6 3.6-2 Rev.5 3.6 3.6-3 Rev.O Tables 3.1-1 and 3.12 N/A Rev.O Table 3.13 N/A Rev 4 Tables 3.1-4 through 3.1-7 N/A Rev.O Tables 3.21 through 3.2-3 N/A Rev.O Tables 3.2 4 through 3.2 5 N/A Rev.1 Table 3.51 Sheet 1 Rev.3 Table 3.5-1 Sheet 2 Rev.4 Table 3.51 Sheets 3 through 8 Rev.3 Tabl: 3.51 Sheets 9 through 15 Rev.4 Revision 5 xxyl
1 LIST OF EFFECTIVB PAGES DEFUl:I ED SAFETY ANALYSIS REPORT Section Effective Pages Revision !
Table 3.51 Sheets 16 through 38 Rev.3 ;
, Figures 3.1-1 through 3.1-19 N/A Rev.0
- Figures 3.21 through 3.214 N/A Rev.O Figures 3.215 through 3.2 22 N/A Rev.4 Figures 3.2-23 through 3.2 30 N/A Rev.O ;
Figure 3.31 through 3.3 3 N/A Rev.5 4.0 4.0-1 Rev.4 i 4.0 4.0-2 Rev.3 4.1 4.1 1 through 4.16 Rev.5 4.2 4.2 1 Rev.5 '
4.2 4.2 2 Rev.4 4.2 4.2-3 Rev.5 4.3 4.31 and 4.3 2 Rev.0 4.3 4.'a 3 Rev.4 4.3 4.3-4 Rev.0 4.3 4.3 5 and 4.3-6 Rev.4 4.4 4.4-1 Rev.4 4.5 4.5 1 Rev.5
'"able 4.1 1, Pages 1 and 2 N/A Rev.5 Tables 4.31 N/A Rev.5 Figure 4.21 N/A Rev.4 Figure 4.2 2 N/A Rev.4 1
5.0 5.0 1 Rev.0 5.1 5.1-1 Rev.0 5.2 5.21 through 5.2 3 Rev.0 5.2 5.2-4 Rev.5 5.2 5.9. 5 and 5.2-6 Rev.0 4 5.2 5.2 7 and 5.2-8 Rev.5 5.3 5.3-1 through 5.3 Rev.0 5.4 5.41 and 5.4 2 Rev.0 5.4 5.4-3 and 5.4-4 Rev.5 5.5 5.5-1 Rev.0 '
5.5 5.5-2 through 5.5-4 Rev.3 l - 5.5 5.5-5 through 5.510 Rev.5 xxvil Revision 5
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LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision 5.6 5.6 1 Rev.1 5.6 5.6 2 Rev.3 5.6 5.6-3 Rev.0 5.6 5.6-4 through 5.6-6 Rev.3 5.6 5.67 Rev.3 5.6 5.6 8 Rev.5 5.6 5.6-9 and 5.610 Rev.3 5.7 5.7 1 Rev.0 5.8 5.8-1 Rev.0 5.9 5.91 through 5.9 2 Rev.0
'I ables 5.5-1 N/A Rev.3 Figure 5.2-1 N/A Rev.5 Figure 5.2-2 N/A Rev.O Figure 5.2 3 N/A Rev.5 Figures 5.2-4 through 5.2-6 N/A Rev.O Figure 5.2 7 N/A Rev.5 Figure 5.31 N/A Rev.O Figure 5.3 2 N/A Rev.5 Figure 5.3 3 N/A Rev.O Figure 5.41 N/A Rev.0 6.0 6.0-1 through 6.0-8 Rev.0 6.1 6.1-1 Rev.5 6.1 6.1-2 through 6.1-4 Rev.0 6.2 6.2-1 through 6.2-3 Rev.0 6.3 6.3 1 Rev 5 6.3 6.3-2 through 6.3-4 Rev.0 6.3 6.3 5 Rev.4 6.3 6.3 6 Rev.0 6.3 6.3 7 Rev.4 6.4 6.4 1 Rev.3 6.4 6.4-2 Rev.4 Tables 6.0-1 through 6.0-5 N/A Rev.O Tables 6.2-1 and 6.2-2 N/A Rev.O Tables 6.3-1 and 6.3 2 N/A Rev. O Figures 6.3-1 through 6.3-6 N/A Rev.0 7.0 7.0-1 Rev.3 7.1 7.1-1 Rev,4 g
Revision 5 xxviii
i i 1IST OF EFFECTIVE PAGES DEFUEl ED SAFETY ANALYSIS REPORT Section Effeslive Paces Revision 7.1 7.1-2 Rev.3 7.1 7.13 and 7.1-4 Rev.4 7.1 7.1 5 Rev.3 7.2 7.21 and 7.2-2 Rev.3 7.2 7.2 3 through 7.2-6 Rev.0 7.3-1 and 7.3 2 Rev.3 7.3 7.3 7.3 3 Rev.0 7.4-1 Rev. 0 7.4 7.5 7.5 1 Rev.4 7.6-1 Rev.3 7.6 8.0 1 Rev. 0 8.0 9.01 Rev. 0 9.0 O
O xxix Revision 5
l.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 ICE as well as by Westinghouse, Bechtel and other organizations. l l
Some documents that are incorporated by reference continue to be updated to assure that 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) 10E 1017. " Security Plan."
(4) PGE 1020, " Report on Design Modi 0 cations for the Trojan Control Building."
(5) PGE 8010, " Nuclear Quality Assurance Program." l (6) POE 1052, " Quality Related List Classification Criteria for the Trojan l Nuclear Plant." ,
(7) 10E 1021, "Offsite Dose Calculation Manual." l (8) PGE 1057, " Trojan Nuclear Plant Certified Fuel llandler Training Program." l (9) PGE-1024, " Trojan Nuclear Plant Security Force Training l .
and Qualification Plan (Defueled Condition)." ,
-(10) PGE-1061, " Decommissioning Plan." l (11) PGE 106 , " Supplement to Applicant's Environmental Report Post Operating l ,
License Stage." ,
i E
L A
V 1.5 1 Revision 5
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I 3 Criterion 61 - EucLSlorage and llandling and Radioactivity Control. The fuel storage and (V
handling, radioactive waste and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.
These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement and filtering systems, (4) with a residual heat removai capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under ac.cident cond'.tions.
The Trojan fuel storage facility has been designed to maintain spent fuel cooling and to prevent criticality as discussed in Section 3.2, Radioactive waste treatment systems are located in the Auxi'iary and Fuel Buildings, which contain or conGne leakage under normal and accident conditions. Adequate shielding is provided as described in Section 5.6.6.1. The enociated ventilation equipment includes filtration which minimizes radioactive material releases associated with normal operation and a postulated spent fuel storage or handling accident. Diverse means are available to supply SFP makeup water for spent fuel cooling.
Criterion 61 is met by Trojan design.
Criterion 62 - Prevention of Criticality in Fuel Storage and IIandling. Criticality in the fuel storage and handling system r. hall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
The restraints and interlocks provided for safe handling and storage of spent fuel at Trojan are discussed in Sections 3.3.1 and 4.2.
The center-to-center distance between tae adjacent spent fuel assemblies in the storage racks is sufficient to ensure a ke rr <0.95, even if unborated water is used to Gil the SFP. The SFP racks are freestanding modules containing a neutron absorber (Boroflex).
(O 3.1-5 l
l
)
I r
l The design of the spent fuel storage rack assembly was such that it was impossible to insert the spent fuel assemblies in other than the prescribed configuration, thereby preventing any l possibility of accidental criticality. One storage rack has been removed from the SFP to facilitate l cleaning of the SFP floor prior to the start ofISFSI loading. The removal of one SFP rack l makes it possible to insert a fuel assembly into the open area that remains. If a fuel assembly is l placed immediately along side a remaining rack, the assembly could be at a spccing less than the l minimum center to-center distance required to ensure against accidental criticality. This l condition was evaluated during the SFP rerack that installed these racks. The prescribed l compensatory measure was to add administrative controls to the fuel handling procedures that l restrict storage of fuel assemblies in the cells af the remaining racks immediately adjacent to the l opening left by removal of the one rack. This ensures that a fuel assembly placed in the open l space will not be adjacent to another fuel assembly. The SFP water is borated 22000 ppm boron to ensure that the worst case configuration, resulting from a dropped fuel assembly, remains subcritical.
O Layout of the fuel handling area is such that the spent fuel casks will not be required to traverse l the SFP during removal of the spent fuel assemblies.
Criterion 62 is met by Trojan design.
Criterion 63 - Monitoring Fuel and Waste Storage. Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.
SFP parameters (level and temperature) and high radiation in the spent fuel storage area are remotely alarmed in the control room. Plant personnel can then take action to modify system lineup or investigate the cause of the abantmal condition.
Criterion 63 is met by Trojan design. g Revision 5 3.1-6
l l
Criterion 64 - Monitoring Radioactivity Releases. Means shall be provided for monitoring
]U the reactor containment atmosphere, spaces containing components for recirculetion of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.
Radiogas and air particulate sampling and measuring equipment is provided to evaluate in-plant conditions under both normal and accident conditions. Measurement capability and teporting of radioactive effluents meet the requirements of Regulatory Guides 1.21 and 4.1.
Criterion 64 is met by Trojan design.
3.1.2 CLASSIFICATION OF STRUCTURES. COMPONENTS AND SYSTiiMS The classification of structures, components and systems for the facility will be per the quality related definition as described in Topical Report PGE-8010, " Nuclear Quality Assurance U Program". Implementation of the classification is described in PGE Report PGE-1052,
" Quality-Related List Classification Criteria for the Trojan Nuclear Plant".
For the permanently defueled condition, the Control, Auxiliary and Fuel Building Complex; SFP, including the spent fuel racks; and the fuel transfer tube are the only structures, systems, or components that are classified as Seismic Category 1 (safety-related). The Containment Structure and the Fuel Building Steel Superstructure are classified as Seismic Category II/I.
3.1.3 WIND AND TORNADO LOADINGS The Trojan Facility is capable of withstanding the effects of severe winds or tornadoes without loss of capability of the safety systems to perform their safety functions. The following sections provide the basis for the design wind and tornado parameters and methods used in meeting the wind and tornado criteria for the Facility.
( $
V 3.1-7 Revision 5 l
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11.3.1 Wind Loadings O
3.1.3.1.1 Design Wind Velocity The design wind velocity for all Category I and Category II structures is 105 mph at 30 ft above the nominal ground elevation of 45 feet MSL. Section 2.3 provides meteorological data for the site, 3.1.3.1.2 Detennination of Applied Forces Shape factors, variation of wind velocity with height, gust factors and methods of converting wind velocities into loads to be resisted by the structures are based on ASCE Paper No. 3269, " Wind Forces on Structures",1961. Table 3.1-1 indicates the dtign loads. These loads are considered in the design of all Category I and Category 11 structures, llowever, since wind and earthquakes are not assumed to act simultaneously, for structures where seismic forces exceed the wind loads, no further analysis is performed. For structures where wind load governs, the load is combined with other appropriate loads as required by the various load equations.
Wind loads are applied to the structures as uniform static loads on the vertical and horizontal projected areas of the structure walls and roof. Roofloads due to wind are treated the same as roof dead and live loads with the direc: ion of loading taken into account. Only dead load is considered as resisting uplift. Ilorizontal wind loads are distributed by the walls to the floor and roof diaphragms which, in turn, transfer the loads to the lateral load carrying elements of the structutes.
3.1.3.2 Tornado Loadings in the continental United States, west of the Rocky Mountains, the occurrence of tornadoes is unlikely. Ilowever, to ensure that any damage which may be sustained by 3.1-8 9
l l
The NRC has reviewed the movement of a transfer and storage cask into the Fuel Building. l The previously existing License Condition (2.C.7), " Spent Fuel Assembly Shipping Cask," l was authorized to be removed by License Amendment Number 196, dated May 19, 1997. l Specific actions to limit floor loadings for hypothetical drops were described in the PGE l License Amendment Applications (LCAs 240 and 237) and are required to be implemented l during cask movements over the affected areas of the Fuel Building by the NRC Safety l Evaluation associated with License Amendment Number 196. The Fuel Building crane is l restricted to operation within a pre-approved envelope as described in Section 4.2.1.
Nuclear fuel storage considerations:
The spent fuel bundles are stored in stainless steel racks in the SFP. The SFP and racks are described in Section 3.2.2.
The reinforced concrete walls of the SFP structure range from approximately 4-feet 6-inches to 6-feet 6-inches, and the base slab has a nominal thickness of 8-feet 4-inches. The concrete has a design strength of 3000 psi. The SFP structure is founded on rock with a design allowable bearing pressure of 20,00u psf. The inside face of the wils and base slab are lined with 1/4-inch-thick stainless steel liner plate to provide leak-tightness. The steel superstmeture of the Fuel Building protects the SFP from the environment.
Exterior walls to Elevation 63 feet of the Fuel Building are precast concrete panels and corrugated metal siding above that. Framing consists of structural steel with reinforced concrete floor slabs. Interior walls are concrete block masonry.
The fuel storage and SFP walls are inherently resistant to missiles. The thermal stresses in the walls of the SFP have been evaluated. No special provisions are made to control cracking of the concrete walls under the thermal stresses mentioned above, but these cracks are minute and p do not extend through the full thickness of the walls. Moreover, these cracks are not unusual LJ 3.2-3 Revision 5
in a reinforced concrete structure. The liner plate on the inside face of the wall and on the floor, due to its ductile nature, will absorb the strain due to the cracking of the concrete and, along with the concrete, will guarantee the leak-tightness of the SFP.
SFP pool drawdown during a tornado has been reviewed and the conclusion is that drawdown is not critical. The capability of the steel superstructure of the Fuel Building to withstand tornado loads is discussed in Section 3.1.3. The crane will not fall off the rails during the postulated tornado.
3.2.1.1.2 Auxiliary Building The Auxiliary Building is a Seismic Category I structure and consists of two Boors below grade (Elevation 45 feet) and four floors above. The portion above grade is structurally connected to the Fuel Building on the east and to the Control Building on the west. The containment is located to the south of the Auxiliary Building. The containment and the Auxiliary Building are separated by a nominal 3-inch expansion joint. According to analysis, this gap is adequate to prevent tha two structures from coming in contact with one another during an earthquake or DBA. The minimum gap during such an event is in excess of 2 inches. A number of framing members of the Auxiliary Building are supported by the Containment wall. These members rest on seats which allow horizontal movement.
Soil interaction between the Auxiliary Building and the adjoining structures during an earthquake has been investigated. Special attention is given to the effect of the containment on the Auxiliary Building. However, since the containment and the Auxiliary Building are founded on rock and the foundation bearing stresses are lower than allowable bearing stresses, interaction for the two structures is not significant.
The exterior walls below grade level (Elevation 45 feet) and all slabs are constructed of reinforced concrete. Interior framing members below grade and all framing members above Revision 5 3.2-4
grade are structural steel. ..erior walls above grade are com..ructed of concrete 'alock masonry at the first floor and metal siding at the top floors. Interior walls are constructed of concrete block masonry.
3.2.1.1.3 Control Building The Control Building is a Seismic Category I structure and consists of four floors above grade (Elevation 45 feet). The Auxiliary Building is located to the east of the Control Building and the two buildings are structurally connected. The Turbine Building, located to the west, is separated from the Control Building by a gap which is adequate to prevent interaction between the two structures during an earthquake.
The framing members for the Cantrol Building are structural steel. Floor slabs are reinforced concrete and precast prestressed concrete panels. Concrete block masonry and reinforced concrete are used for walls.
3.2.1.1.4 Modification to the Complex The Control Building was modified during 1980 and 1981 to increase the seismic capability of the overall Complex. The modification program was developed with two basic objectives:
first, to add strength to the building; second, to strengthen the connections among certain wall panels to achieve a better group action and improve overall seismic performance.
Strengthening of the Control Building was done by the addition of four new structural elements: three parallel walls running in the north-south direction and a steel plate added to one wall. Two of the new walls closed the previous railroad bay openings in the east and west walls on column lines N and R, and the third wall is an interior shear wall crossing the previous railroad bay on column line N'. The upper portion of the west wall of the Control Building along column line R was further stiffened and strengthened by the addition of a f])
3.2-5 Revision 5
l 3-inch-thick steel plate. The previous railroad bay in the Control Building was thus totally enclosed, and the railroad spur that used to run through the Control Building was terminated in the Turbine Building.
In addition to the four new structural elements, structural improvements were also made at several locations to strengthen connections. Four structural improvements involved strengthening the existing bolted beam-column connections of coh?mn 46-N beneath Elevation 61 feet, Elevation 77 feet and Elevation 93 feet, and column 46-R beneath Elevation 77 feet by welding additional plates. The remaining structural improvements consisted cf connecting the horizontal reinforcing steel at the following walls to make the steel continuous:
(1) The 41 wall at column 41-Q, Elevation 45 feet to Elevation 65 feet .
(2) The 46 wall at column 46-N', Elevition 45 feet to Elevation 61 feet (3) The 55 wall at column 55-Q, Elevation 45 feet to Elevation 61 feet (4) The 55 wall at column 55-N', Elevation 45 feet to Elevation 61 feet.
Figures 3.2-10 through 3.2-14 show the stmetural modifications to the Control Building.
3.2.1.2 Design Bases The design bases for the Complex are defined within the following classifications:
(1) Support - in addition to their own weight, the stmetures are designed to supp :rt the reactions and weight of the components of systems shown in Figures 3.2-15 through 3.2-22. Figures 3.2-15 through 3.2-22 reflect the structures and Revision 5 3.2-6
/
(%) components of systems that existed during Trojan Plant operation, and are included in the DSAR for general information. The components of systems and the structures shown on the figures are being de-activated, decommissioned, or removed in conformance with the Trojan Nuclear Plant Decommissioning Plan, PGE 1061, which was approved by the NRC, by letter dated April 15, 1996.
These figures are, therefore, historical in nature and are not required to be updated as the decommissioning activities are completed. Control of formal engineering drawings is maintained in accordance with PGE-1061.
(2) Diological shield - the shielding design teatures are based on potential radiation sources during normal Plant operation, shutdown, and energency operations (0 The shielding design conservatively bounds the current defueled state.
(3) Missile shield - the missile shielding design bases are given in Section 3.1.5.
fi d
(4) Administrative control of access - the design bases are given in Section 7.7.
(5) Atmosphere control - the design bases include structural leak-tightness. external and internal pressures from usage of ventilation systems, and equipment shelter from the elements (D.
(6) Liquid control - protection from the efflux of water is required for the SFP and in areas where liquid radioactive wastes are handled by radioactive waste treatment systems (0, (7) Seismic protection - design bases are given in Section 3.1.6.
(8) Wind protection - design bases are given in Section 3.1.3.
,Q 3.2-7 Revision 5
3.2.1.3 Aeolicable Codes. Standards. and Soecifications 3.2.1.3.1 Applicable to Original Prctedure This section lists the codes, specifications, regulatory guides, and other documents used in the structural design of the stmetures listed in this section except for the modifications of the Complex and the Fuel Building steel superstructure. Section 3.2.1.3.2 lists the codes, specifications, regulatory guides, and other documents used in the design of the modifications to the Complex. Section 3.2.1.3.3 lists the codes, specifications, and other documents used in the assessment of the Fuel Building superstructure.
(1) American_ Concrete Institute (ACI):
ACI 315-65 " Manual of Standard Practice for Detailing Reinforced Concrete Standards".
ACI 318-63 " Building Code Requirements for Reinforced Concrete".
(2) American Institute of Steel Construction (AISO, " Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings", Sixth Edition,1963.
(3) American Society of Mechanical Engineen, "ASME Boiler and Pressure Vessel Code", Sections III, V, VIII, and IX,1968 Editions.
(4) Uniform Building Code (UBC),1967 Edition.
(5) AEC Publication TID-7024, " Nuclear Reactors and Earthquakes".
(6) American Society of Civil Engineers (ASCE), " Paper No. 3269, Wind Forces on Stmetures".
Revision 4 3.2-8 !
s i-
- describid in Section 3.2.1.4 are applicable to the design of foundations. Design and analysis ,
procedures applicable to foundations are described in Section 3.2.1.5. Criteria for foundations are the same as those listed for the stmetures in general (see Section 3.2.1.6). Description of the reinforced concrete and structural steel used in the constmetion of foundations is in Section 3.2.1.7 Floor slabs were coated to prevent possible absorption of contaminated fluids and to facilitate decontamination.
l 3.2.2 SPENT FUFI POOL AND FUEL STORAGE RACKS The SFP is the storage space for irradiated fuel from the reactor. -
- 3 2.2;l Design Bases The original design bases for the SFP and rack is designed for the following:co,11.12)
- (1) Store 1408 fuel assemblies with maximum enrichmen 4.5 wt% and an average region burnup of 55,000 mwd /MTU. Some of the 1408 storage cells are used to store fuel rod
! storage canisters, radioactive filters, debris or specimen assemblies associated with refueling evolutions. Note: one SFP rack consisting of 121 cells has been removed l leaving 1287 cells, l C
~ (2) Maintzin kerr s0.95 during normal conditions with SFP water between 40*F and 140*F
. and a boron concentration of 0 ppm (demineralized water).
1
-(3) Maintain k rre 50.95 for abnormal conditions, including:
~
~ (a) L6ss of normal SFP cooling (water temperature of 212*F at SFP surface).
O 3.2-35 Revision 5 a
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i (b) Drop accidents involving items which may be transferred or handled over the fuel racks. These items are shown in Table 3.1-4.
(c) Accidental drop of a fuel assembly in any position or orientation.
(d) Effects on fuel assemblies resulting from an earthquake or missile (See Table 3.1-4).
During abnormal events, a soluble boron concentration of 2000 ppm is assumed.
Variations in rack dimensions, neutron absorber parameters, fuel parameters, and fuel location permitted by fabrication tolerances are included in the analysis. Calculations were performed to determine the sensitivity of ke rt to abnormal SFP water temperatures which may be encountered during a loss of cooling water incident.
(4) Maintain fuel cladding integrity in the event forced cooling is lost and cooling occurs by SFP boiling (212"F) at the water surface. Evaporative losses are made up by a variety of sources as discussed in Section 4.3.1.
(5) The structural design of the spent fuel racks is in accordance with NRC Standard Review Plan 3.8.4 Appendix D, " Technical Position on Spent Fuel Racks." Construction materials for the racks conform to the requirements of ASME Section III, Division 1, Subsection NF.
The governing code for rack design and analysis is ASME Section III, Division 1 Subsection NF for Class 3 component supports. The design loads are specified in Reference 10.
(6) The entire structure, including the spent fuel racks, has been designed to Seismic Category I requirements.
Revision 5 3.2-36
/ (7) The dose rates at the surface of the SFP from spent fuel assemblies do not exceed 2.5 mrem /hr during spent fuel transfer and storage. Dose rates at the outside surface of the walls of the SFP do not exceed the maximum radiation zone level for the aredl) .
(8) A f.4 handling accident involving dropping of a single spent fuel assembly in the SFP from its maximum attainable height will not result in offsite radiation doses to the public exceeding the values calculated in Section 6,2.
l 3.2.2.2 System Design The SFP is a reinforced concrete structure with seam-welded stainless steel plate liners. The pool 2
volume is approximately 51,900 ft3 with a surface area of 1300 ft . The pool is filled with borated water which is maintained at a concentration of 22000 ppm boron.
p The spent fuel assemblies are stored in storage racks in parallel rows having a center-to-center U distame of 10.5 inches in both horizontal directions. 'line racks are freestanding modules containing a neutron absorber (Boroflex). Burnable poison rod assemblies, neutron-source assemblies and thimble-plugging devices removed from the reactor are also stored in the SFP.
Adjacent to the SFP are two small pools. One is the fuel transfer canal which is connected to the refueling cavity (inside the containment) by the fuel transfer tube. The other is the spent fuel cask loading pit. Leak-tight doors have been provided between the SFP and these two smaller pools to allow underwater movement of the assemblies between pools. The doors open into the SFP so that when the doors are closed with the adjacc.nt pools drained, water pressure tends to seal the doors.
Addition.nlly, each door is equipped with an inflatable boot seal around its periphery which is inflated when the door is closed using the instrument and service air system.
The water level in the SFP is maintained to provide at least 23 feet of water above the top of a spent fuel assembly in the storage racks, and at least 9.5 feet above the active portion of the fuel assembly v
3.2-37 Revision 5
during fuel transfer operations. This water barrier serves as a radiation shield, enabling the gamma dose rate at the pool surface from the spent fuel assembly to be maintained at or below 2.5 mrem /hr.
Overflows from the SFP drain into the SFP ventilation system (AB-4) exhaust ductwork and are directed to the dirty waste drain tank.
Beneath the SFP liner is a network of monitoring trenches which will collect any leakage through the liner. The trenches drain through normally open valves to the dirty radioactive waste treatment system. The leak detection valves are arranged into manifolds that are inspected periodically to monitor for SFP liner leakage.
Ventilation systems remove gaseous radioactivity from the atmosphere above the SFP and discharge through the Plant vent. The ventilation systems are described in Section 5.2. These ventilation systems are monitored for radioactivity by Process and Effluent Radiation Monitoring Systems (PERMS) which are described in Section 5.5.
A SFP Area Radiation Monitoring System (ARMS) is provided for personnel protection and general surveillance of the SFP area. These ARMS are discussed in Section 5.6.1.
The SFP water chemistry is sampled and maintained in accordance with Section 4.1.2. The environment to which the SFP liner and spent fuel are exposed is not conducive to corrosion.
3.2.2.3 Design Evaluation The center-to< enter distance between the adjacent spent fuel assemblies and the fixed neutron absorber are sufficient to ensure a ke rrs0.95 even if unborated water and fresh nondepleted fuel, enriched in U-235 to 4.5 wt%, are in the SFP.
O Revision 5 3.2-38
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The design of the spent fuel storage rack assembly was such that it was impossible to insert the spent l fuel assemblies in other than prescribed locations, thereby preventing any possibility of accidental criticality. One storage rack has been removed from the SFP to facilitate cleaning of the SFP floor l prior to the start ofISFSI loading. The removal of one SFP rack makes it possible to insert a fuel- l assembly into the open area that remains. If a fuel assembly is placed immediately along side a l remaining rack, the assembly could be at a spacing less than the minimum center to-center distance l required to ensure against accidental criticality. This condition was evaluated during the SFP rerack - l that installed these racks. The prescribed compensatory measure was to add administrative controls to l 1 the fuel handling procedures that restrict storage of fuel assemblies in the cells of the remaining racks l immediately adjacent to the opening left by renoval of the one rack. This ensures that a fuel assembly [
placed in the open space will not be adjacent to another fuel assembly. l 2 -
Mechanicsl and electrical stops are provided on the Fuel Building bridge crane rails to prevent an
' inadvertent traverse of the pool with heavy loads.
The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced SFP cooling due to failures
- of the non Seismic Category 1 piping and the fuel transfer doors and determined that the loss of SFP
- cooling and inventory would not lead to uncovering of the fuel. Sufficient time would be available to establish makeup flow to the SFP.
Lines entering the SFP which could siphon the pool to Elevation 76-feet 7-inches or below (approximately 10 feet above fuel elements) are equipped with siphon breakers e Elevation 83-feet I l-
-inches.
l(3 3.2-39 Revision 5
i I
1.2.2.4 Tests and Inspections Neutron absorber coupons arc located in the SFP and used for periodic verification of absorber effectiveness. The verification surveillance is every 4 years. Coupon inspection includes visual inspection, hardness, weight, and dimensional measurements.
The SFP liner leak detection manifold is inspected periodically to monitor for SFP liner leakage.
3.2.2.5 Instrumentation Application A level switch is provided in the SFP for the purpose of annunciating high and low water levels to the control room. Local level indication is installed in the SFP.
SFP water temperature indication and a high temperature alarm on the sequence of events recorder are provided in the control room.
O 3.2.2.6 SFP Stmeture Re-evaluation for Beyond Design Basis Seismic Motions The SFP stmeture was evaluated for ground motions that would result from thc xcurrence of a conservatively postulated Seismic Margin Earthquake (SME).
Cascadia Subduction Zone carthquake source interface events in the magnitude range of M 8 to 9 and intrastab events in the range of M 7 to 7 in, and random crustal earthquakes in the range of M 6 to 6 la were considered to envelop hypothetical earthquake ground motion exposures to the Trojan site.
The bounding case SME was determined to be a Cascadia Subduction Zone intraslab earthquake of magnitude range M 7 to 7 in at a hypocentral distance (source to site distance) of between 55 and 60 km. The corresponding peak horizontal ground acceleration at Trojan would be 0.38 g.
O Revision 5 3.2-40
~
' The SFP structure was deterministically found to have large to very lage capacity margins over SME demands. : Peak horizontal ground accelerations on the order of 2 g would be required to -
apprcsch the threshold of unacceptable damage to the SFP walls. There are no rational earthquake sources that could produce anywhere near this level of ground acceleration at the Trojan site. There is,' therefore, a very high level of confidence that the SFP structure would not be damaged by the .
postulated occurrence of a SME.
A probabilistic assessment was also performed that concluded that the annual probability of
- unacceptable SFP wall structural performance due to an carthquake producing several times the SME
' level of ground acceleration would be considerably less than 5x10s, I
b i
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3.2-41 Revision 5 n~ No -
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,f] 3.3 AUXILIARY SYSTEMS v -
This section discusses the auxiliary systems that are used to support the storag- of spent fuel at Trojan. This section includes discussions on the fuel handling system, SFP cooling and demineralizer system, component cooling water system, service water system, compressed air system, makeup water treatment system, equipment and floor drain systems, Plant discharge and dilution structure, 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.
3.3.1 FUEL IIANDLING SYSTEM 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. The transfer of fuel to the Containment
\ 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 or jamming of fuel assemblies during transfer operation,.
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3.3-1 Revision 5
(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 hft height which is determined by the length of the long-handled tool, so that the mindmum 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 Descriotion 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 ::hielding is maintained. Fuel assembly inserts, such as thimble 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 Building bridge 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 25 tons. The crane and accessories have been designed and constructed for indoor service and were designed to handle new and spent fuel containers between the railroad cars and loading and unloading pits. Movement of 9
3.3-2
. . _~ . - ._- _
i handling radioactively contaminated fluids and 'he primary heat sink (SWS) With the reactor permanently defueled, the CCWS provides the required cooling water for removal of heat from the SFP via the SFP cooling water heat exchangers. The CCWS may also remove heat from the l SFP to Containment via the Containment Air Coolers (CACs). l With the reactor permanently defueled, the CCWS no longer performs any safety functions.
3.3.3.1 Design Bases Each loop of the CCWS was originally designed for design basis accident heat loads and normal reactor operation heat loads, including the heat loads from both SFP cooling water heat exchangers (1.3 x 107 Bru/hr) m. The current heat load from the SFP cooling water heat exchangers is a small fraction of the original design basis CCWS heat load ensuring that a single CCWS loop can provide the required heat removal capacity.
3.3.3.2 System Descriotion The CCWS is shown in Figure 3.3 2. The components included in the system are:
4 (1) - Three CCWS pumps (P-210A, D, and C). l (2) Two CCWS heat exchangers (E-204A and B).
(3) Two CCWS surge tanks (T-204A and B).
(4) One chemical additive tank (T-209).
(5) Two CCWS makeup pumps (P-218A and B).
(6) Interconnecting piping, valves, and instrumentation.
- The CCWS is a closed-cycle cooling system with the fluid being continuously recirculated through the system by the CCWS pumps. No part of the system is normally open to the 3.3-9 Revision 3 n w
atmosphere. Ileat is removed from the system by the flow of service water through the tube side of the CCWS heat exchangers.
O Three pumps provide the required circulation through the system. Pumps P-210A and P-210 C are identical. Pump P-210D was sized for current heat loadi and is smaller than the other two l 1
pumps. Pump P-210A is connected directly to loop A and pump P-210D !s connected directly to l loop B. Pump P-210C can be connected to either loop to replace either F-210A or P-210D by opening normally locked-closed valves. Power to P-210A is supplied from 4160-V bus Al and P-210D is supplied from 480-V MCC bus B 35. Power to P-210C can be supplied from either bus Al or A2 through a manually operated transfer switch.
Two identical CCWS heat exchangers provide the required heat transfer from the CCWS to the SWS. The service water, which has a greater tendency to foul heat exchanger surfaces, circulates through the heat exchanger tubes to facilitate cleaning of the water surfaces. The CCWS heat exchangers are designed to provide the required heat removal capacity assuming a maximum service water temperature of 75*F.
O A surge tank is provided in each of the CCWS loops. The primary function of the tanks is to provided a static pressure on te CCWS sufficient to maintain the fluid pressure throughout the system above the fluid vapor pressure. The tanks also provide the following functions: (a) a means for damping trans'ent pressures, (b) a means of monitoring the vol.ime of fluid in the system, (c) 1 means of providing for expansion and contraction of the fluid in the system, (d) a source of pressure relief to the system through the safety relief valves on the tanks to the dirty radioactive waste treatment system (DRWS).
A cushion of nitrogen is used to provide the required pressure in the CCWS surge tanks. The l nitrogen is normally supplied from the Plant ni.rogen storage system. Self-contained pressure-regulating valves are installed in the nitrogen supply lines between the nitrogen sources and the surge tanks. The pressure may also be adjusted manually using the solenoid valves installed in Revision 5 3.3-10
(9 V
the nitrogen supply lines from the Plant nitrogen storage system. Each tank is equipped with two safety relief valves to prevent overpressurizing the sys,em.
Provisions exist for adding corrosion inhibitor to the system.
Two identical CCWS makeup pumps are installed to furnish makeup water to the system. Normal nakeup water is supplied f.om the deminerativd water storage tank using the demineralized water transfer pumps.
For the defueled condition, train independence and automatic isolation of selected loads are not required. A single CCWS pump and a single CCWS heat exchanger provide excess heat removal capacity for the current heat loads.
Piping in the portions of the CCWS providing SFP cooling is sea : carbon steel, fabricated
.A and installed in accordance with the requirements of ANSI B.31.1.0, Code for Power Piping.
(
3.3.3.3 Design Evaluation i
The CCWS does not perform any safety functions. Loss of component cooling water to the SFP cooling water heat exchangers will cause the SFP water temperature to slowly rise. If the component cooling water cannot be restored to the heat exchangers, then the SFP water temperature will continue to rise, increasing the evaporation rate and possibly resulting in boiling wi+' he SFP. The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to the SFP a.. , detemiined that 2.Ticient time exists to effect repairs to the cooling system or to establish makeup flow prior to uncovering the spent fuel. Makeup water is available from a varier, of sources as described in Section 4.3.1.
3.3-11 Revision 5
3.3.4 SERVICE WATER SYSTEM The SWS, shown in Figure 3.3-3, is designed to provide water from the Columbia River to cool equipment and to supply water to various systems and equipment. With the reactor permanently defueled, the primary functions of the SWS are reduced to providing cooling water to the l component cooling water heat exchangers and selected room cooler uma as well as provide makeup to the spent fuel pool.
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 defueled, system design requirements are reduced substantially from the original design bases of the system.
Ileat transfer equipment was selected based on a temperature of 75 F, which exceeds the highest recorded Columbia River water temperature.
l The system design includes provisions for inhibiting long-term corrosion and organic fouling of the system water passages.
O Revision 4 3.3-12
-- . - . . . . . . ..-- . - . .- - . . - ~ . - . . ~ . - . - - .- -.---
i l
i .
l
'3 .3.4.2 System Description q O :
The SWS is shown in Figure 3.3-3. The components included in the system are:
(1) Three SWS pumps (P-108A, B, and C).
4
'(2) Two SWS strainers (F-101 A and B).
(3) Two SWS booster pumps (P-148B and D). l
+
J U (4) Interconnecting piping, valves, and instrumentation.
l Water is supplied through the trash rack and traveling water screens at the intake structure. The water entering the system is periodically chlorinated for microbiological control.
One traveling water screen is installed in the flow path to each of the two independent flow paths to the service water pumps at the intake stnicture. The screens are automatically cleaned by the screen m.a system which consists of two vertical pumps taking suction on the downstream side i
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 acrass the screen.
Three identical service water pumps take suction from the river through the traveling water screens at the intake structure. Each pump is designed to provide 100 percent of the flow -
requirements. Pump P-108A is aligned to loop A and pump P-108B is aligned to loop B.- Pump.
4 P-108C can be aligned to either loop by opening normally locked-closed valves. - Power to P-108A is supplied from 4160-V bus Al and P-108B is supplied from 4160-V bus A2 Power to P-~
, 108C can be supplied from either bus Al or A2 through a manually operated transfer switch.
qi x) -
3.3-13 Revision 5 s
Lubricating water is supplied to the three service water pumps. River water can also be supplied to the SWS using pumps P-167A and P-16713.
Two identical service water strainers are provided for straining the discharge from the service water pumps. One strainer is located in the sapply header to each loop.
l Two identical service water booster pumps, P-148B and D, are provided to boost the pressure in the water supply lines to components served by the system except the supplies to the CCWS heat exchangers, makeup water to the CCWS, bearing tube water and SWS strainer backwash. These l are supplied directly from the service water pumps. Only one service water booster pump is required to provide 100 percent of the loop flow requirements.
l A source of SFP makeup water is provided from the service water booster pumps.
For the defueled condition, train independence and automatic isolation of the Seismic Category II l loads are not required. A single SW loop provides excess capacity for the current loads.
Discharged water is dechlorinated at the Plant discharge and dilution structure before being discharged to the Columbia River as discussed in Section 3.3.9.
3.3.4.3 Design EvaluMicB ;
During normal operation, the only requirement for the SWS is to maintain the SFP temperature sl40*F, Loss of the SWS will cause the SFP water temperature to slowly rise due to loss of heat removal capability from the SFP cooling water heat exchangers (due to loss of component cooling water cooling). The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP abow the spent fuel elements. Section 6.3 analyzed the loss of forced O
Revision 5 3.3-14
/ cooling to the SFP and determined that sufficient time exists to effect repairs to the cooling
(]
system or to establish makeup flow from an alternate makeup source prior to uncovering the spent fuel.
3.3.5 COMPRESSED AIR SYSTEM The compressed air system provides the Plant compressed air requirements for pneumatic instruments and valves and for service air outlets located throughout the Plant which are used for operation of pneumatic tools. The system does not perform any safety functions.
The system uses water-cooled aftercoolers and compressors. The air receivers are connected to a common compressed air header which connects to the air filter unit. The discharge of the air filter unit connects to the air-dryer unit inlet and the service air header. The instmment air header is connected to the air-dryer unit discharge. Each air header supplies branch lines which supply g instrument air and service air to the required loads throughout the Plant. The instrument and U service air system provides air to the inflatable seals for the SFP gates and to the CCWS air-operated isolation valves (CV-3303, CV-3287, CV-3304, and CV-3288). Loss of instrument air to the CCWS isolation valves would cause them to fait closed causing a loss of forced cooling to the SFP (due to loss of component cooling water to the SFP cooling water heat exchangers). The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to the SFP and determined that sufficient time exists to effect repairs to the cooling system or to establish makeup flow prior to uncovering the spent fuel.
3.3.6 BORIC ACID BATCil 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.
]
3.3-15 Revision 5
3.3.7 MAKEUP WATER TREATMENT SYSTEM The Makeup Water Treatment System provides demineralized water of the required quality to meet Plant needs. Makeup water is processed and then stored in the demineralized water storage tank where it is available to meet Plant needs. The DWST is a source of SFP makeup water. The water is transferred to the SFP using a demineralized water transfer pump.
3.3.8 EOUIPMENT AND FLOOR DRAIN SYSTEMS The following equipment and 1100. drainage systems are provided for the Plant:
(1) Dirty Radioactive Waste Treatment System (DRWS) drains.
(2) Clean Radioactive Waste Treatment System (CRWS) drains.
(3) Oily waste system.
(4) Acid waste system.
(5) Sanitary waste system.
The equipment and floor drain systems do not perform any safety functions.
The DRWS and CRWS are designed to collect liquid waste from areas containing equipment that handles radioactive fluids. These systems are designed to control the spread and release of radioactive particulates by directing potentially radioactive fluide to the radioactive waste treatment systems. These systems are described in Section 5.3.
O Revision 5 3.3-16
i I
3 J
- / The oily waste system collects the waste from floor and equipment drains in places where the a
4 waste is not potentially radioactive. The waste _is conveyed to a settling tank where the oil is 3 separated prior to releasing the water to the Plant discharge and dilution stmnure.
x c 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 prcividing a surface resistant to corrosion. The waste is drained into the acutralizing 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.
. The sanitary system collects waste from floor drains in toilet rooms, shower rooms not requiring ;
radioactive waste connecilons, and the plumbing fixtures.
I 3.3.9 PLANT DISCHARGE AND DILUTION STRUCTURE
, The Plant discharge and dilution structure receives the Plant liquid radioactive and chemical i 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 effluent. Diluted chemical waste discharge concentrations '
are diluted to meet the requireinents of the State of Oregon Department of Environmental Quality.
L The structure does not perform any safety functions.
3.3-17 Revision 5
. ,, ., , . - . - . - _ . - .. .= . - .a .-. --
l 3.3.10 PRIMARY SAMPLING SYSTFAI The primary sampling system (PSS) is designed to collect samples from various Plant systems for chemical and radiochemical analysis. Sample points include:
(1) SFP cooling water heet exchanger.
(2) SFP - after demineralizer.
(3) Cler.n waste receiver tank.
(4) Treated waste monitor tank.
(5) Liquid radwaste demineralizer effluent.
The samples are routed to a common location in the radiation sample room at Elevation 45 feet in the Auxiliary Building. The grab sample points are located under a forced ventilation exhaust hood which exhausts i e the vent collection header. Any liquid leakage is collected in the sample sink which drains to the clean waste receiver tank. The PSS operating procedures require that sufficienf surge and sample volume be drawn to ensure that the sample is representative of the process.
3.3.11 FIRE PROTECTION SYSTEM ANDS.ROGRAM Fire Protection is provided for the Plant as described in the Topical Report, " Trojan Nuclear Plant Fire Protection Program", PGE-1012.
O Revision 5 3.3 18
(~'T 3.3.12 CONTROL IU)OM ll ABITABILITY V
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 irem the control roorn to testore 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.
LL13 SEIShilC_ INSTRUMENTATION Seismic instrumentation for the facility consists of a multielement seismoscope, acceleration time-O history devices and peak recording devices. The multielement seismuscope is rigidly attached to the containment base slab. The acceleration time history recording is done by sensors at appropriate !ocations in the facility. Peak acceleration recorders are also installed at appropriate locations in the facility. The seismic instmmentation satisfies 10 CFR 100, Appendix A which requires instrumentation so that the seismic response of features important to safety can be determined promptly to permit comparison of such response with that used as the design basis.
,3
( )
v 3.3 19 Revision 5
1
3.6 REFERENCES
REFERENCES FOR SECTION 3.1
- 1. Trojan Nuclem Plant. Final Safety Analysis RepoII, Amendment 19 (December 1992).
- 2. BC-TOP-4, Scismic Analyses of Stmstures and Eauipment for Nudear Power Plants, Bechtel Corporation (April 30,1971).
- 3. BC-TOP-4A, Scismic Analyses of Stmetures and Equipment for Nuclear Power Plants, Revision 3, Bechtel Power Corporation (November,1974).
- 4. PGE-1020 " Report on Design hiodifications for the Trojan Control Building"," Revision 4, Portland General Electric Company (February 12, 1980).
- 5. I, hi Idriss, Ground Motions. Trojanhuslear Plant. Rainier. Oregon (November 1970).
6, Sdsmic Design Criteria for Nuclear Power. Plants, Newmark End llall, as presented at the Fourth World Seismic Conference held at Santiago, Chile (1969).
O 7. R. A. Parmelee, lluilding Foundation Interaction Effes;, Eng. hiech. Div. Spec Conf. ASCE V
(October 12, 1966).
- 8. E. L. Wilson, A. Der Keureghlan, and E. P. Bayo, "A Replacement for SRSS hiethod in Seismic Analysis", Earthquake Engineering and Structural Dynamics, Volume 9,1981.
- 9. PGE-iOS2, " Quality Related List Classification Criteria for the Trojar. Nuclear Plant,"
Amendment 2 (August 1993).
REFERENCES FOR SECTION 3.2
- 1. Itojan Nuclear Plant. Final Safety Analysis Report, Amendment 19 (December 1992).
- 2. J. K. hicCall, G. Ferrell, and J. V. Rote, Alicisment of Containment Structure Post-Icusioning System Surveillance Program, Bechtel Power Corporation (hiay 1973).
pd 3.6 1
1
- 3. PGE-1020, ~ Report on Design hiodifications for the Trojan Control Building", Revision 4 g (February 12, 1980). W
- 4. PGE letter to A. Schwencer dated June 29,1979, Response to NRC Staff Question 11(a) dated hiay 18, 1979.
- 5. hi. llatzinikolas, J. Longworth, and J. Warwaruk, " Evaluation of Tensile Bond and Shear Bond of blasonry by Means of Centrifugal Force". Alberta hiasonry Institute, Edmonton, Alberta.
- 6. P. A. Ilidalgo, R. L. hiayes, II. D. hicNiven, and R. W. Clough, " Cyclic Loading Tests of hiasonry Single Piers - Volume 3" - CD/EERC 79 ll2, College of Engineering, University of California, Berkeley, California, hiay,1979.
- 7. S. J. Chen, P. A. Ilidalgo, R. L. hiayes, R.W. Cough, and II. D. McNiven, " Cyclic leading Tests of hiasonry Single Piers - Volume 2" - lleight to Width Ratio of 1,".Repon Number UCH/EERC-78-28, College of Engineering, University of California, Berkeley, California, December,1978.
- 8. "Bechtel Quality Assurance Program for Nuclear Power Plants", BQ-TOP-1, Revision 2A, Bechtel Engineering Corporation (July 1977).
- 9. " Licensee's Testimony on hiatters Other thaa Structural Adequacy of the hiodified Complex", prepared by Brochl, et.al., submitted h1 arch 17,1980.
- 10. PGE-1037, " Trojan Nuclear Plant Spent Fuel Storage Rack Replacement Report". (July
] 1983). Illstorical only.
l !!. Trojan Operating License, No. NPF-1, dated November 21,1975. Revised by License l Amendment No.196, Dated hiay 19,1997,
- 12. Letter, B. D. Withers (PGE) to L Frank (OJOE), dated April 17, 1984.
REFERENCES FOR SECTION 3.3
- 1. Isolan Nuclear Plant. Final Safety Analysis Reppa, Amendment 19 (December 1992).
- 2. American.Huclear SociegStandard ANSl/ANS 5.1-1979,
- Decay lleat Power in Nuclear Reactors", approved by ANS 5.
O Revision 5 3.6-2 I
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l 9 4.1 OPERATION DESCRIPTION Operational activities of the defueled Plant are primarily involved with maintaining the spent fuel cooling and supporting systems. Spent fuel pool (SFP) cooling is normally provided by a closed loop system consisting of two pumps and two heat exchangers with a common suction and discharge flow path, in the event forced SFP cooling is lost, the only requirement to assure adequate cooling is to maintain SFP level so that the spent fuel elements are not exposed. Section 4.3 discusses the operation of SFP cooling and support systems.
4.1.1 CRITICAL.lTY PREVENTION The prevention of criticality is primarily addressed by the design of the SFP including the storage racks. A discussion of the design bases is conta;ned in Section 3.2.2. The design of the SFP is such that criticality prevention is provided by maintaining center to-center distance between adjacent fuel assemblics and the fixed neutron absorber contained in tne spent fuel storage racks. The design of the spent fuel storage rack assembly was such that it was l impossible to insert spent fuel assemblies in other tnan prescribed locations, thereby maintaining the required separation and preventing accidental criticality. One storage rack has l been removed from the SFP floor prior to the start of ISFSI loading. The removal of one SFP l rack makes it possible to insert a fuel assembly into the open area that remains. If a fuel l assembly is placed immediately along sid: a remaining rack, the assembly could be at a l spacing less than the stinimum center to-center distance required to ensure against accidental l criticality. This condition was evaluated during the SFP rerack that installed these racks. The l prescribed compensatory measure was to add administrative controls to the fuel handling l procedures that restrict storage of fuel assemblies in the cells of the remaining racks l immediately adjacent to the opening left by removal of the one rack. This ensures that a fuel l assembly placed in the open space will not be adjacent to another fuel assembly. The design l assumes that all storage locations contain fuel umblies with a maximum enrichment of 4.5
( percent (U 235), Credit is not ttken for boration of the SFP cooling water for bulk 4.1-1 Revision 5 l
l I
temperatures between 40*F and 140*F. A soluble boron concentration of 2000 ppm was assumed for abnormal conditions such as:
(1) Loss of SFP cooling (water temperature of 212'F at the SFP surface).
(2) Drop accidents involving items that may be transferted or handled over the fuel racks.
(3) Accidental drop of a fuel assembly in any position or orientation.
(4) Effects on fuel assemblies resulting from an earthquake or tornado missile.
To ensure design requirements are satisfied, administrative controls and operating procedures are employed.
OI Plant operating procedures direct SFP temperature to be maintained between 40*F and 140'F and SFP boron concentration to be.2.2000 ppm. Procedures also limit movement of loads over the spent fuel that could result in unpact energies > 240,000 in.-lbs. This load limit provides assurance that in the event of a fuel handling accident, no more than the contents of one fuel assembly will be mptured as analyzed in Section 6.2.4. To minimize the probability of a fuel handling accident, fuel handling operations are performed in accordance with approved Plant procedures under the direct supervision of a Certified Fuel llandler (CFH).
Lil CllEMISTRY CONTROL To minimize corrosion and contamination, chemistry of the SFP cooling system is maintained.
Table 4.31 provides a listing of chemistry parameters monitored and their frequency.
O Revision 5 4.1-2 !
k Corrective measures for out of specification chemistry will be taken upon discovery. The spent fuel purification subsystem is used to maintain satisfactory S?P purity and clarity. The {
capability for boric acid addition is required to ensure SFP coolant can be maintained l r
>_2000 ppm. Boric acid normally will be mixed in the boric acid batch tank and gravity-drained to the SFP.
4.1.3 INSTRUMENTATION The primary instrumentation associated with operation of the Plant is associated with the SFP cooling system. This instrumentation provides the operating staff with indication of SFP level, pump suction and discharge pressure, and temperatures throughout the system.
Alarms are provided to the control room for abnormal SFP level or high SFP temperature.
There are no automatic actions performed by SFP cooling instrumentation with system operation twing manually controlled. A more complete description of the SFP cooling system s
) is provided in Section 4.3.
4.1.3.1 Seismic Monitoring Instrumentation The seismic monitoring instrumentation shown in Table 4.1 1 shall be operable. With one or more of the seismic monitoring instruments inoperable, the inoperable instrument (s) shall be restored to operable status within 30 days.
With one or more seismic monitoring instmments inoperable for more than 30 days, prepare and submit a report to the NRC within the next ten days outlining the cause of the malfunction and the plans for restoring the instmment(s) to operable status.
Each of the seismic monitoting instmments shown in Table 4.1 1 shall be demonstrated operable by performance of a channel check, channel calibration, and channel functional test at the frequencies listed in the table. These surveillance activities shall be performed within the 4.1-3 Revision 5
l specined laterval with a maximum allowable extension not to exceed 25 percent of the ej specif.al interval.
Each of the seismic monitoring instruments shown in Table 4.1-1 that is actuated during a seistnic : vent shall be restored to operable status and have a channel calibration performed within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. tollowing a seismic event. Data shall be retrieved from actuated instrunients and 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, portions of the Component ;
Cooling Water System (CCWS) and Service Water System (SWS) necessary to support SFP cooling, and components that support the emergency plan, Ore protection plan, security plan or other licensed condition.
4.1.5 ADMINISTRATIVE CONTROL OF SYSTEMS Plant systems and components that are no longer required to support defueled Plant operations will be deactivated. To prevent deactivated systems or components from affecting systems that are required to support defueled Plant operations, the following guidelines have been incorporated into Plant procedures:
(1) Those systems required to support defueled Plant operations shall have physical isolation from systems or components that are deactivated. Isolation will be incorporated into Plant operating procedures.
Revision $ 4.1-4
(2) Deactivated systems will be de-energized and isolation boundaries established to provide appropriate facility aad personnel safety.
(3) Where practical, combustible material such as charcoal from filters, lube oils, electrohydraulic control fluids, etc., should be removed from systems or components not required for defueled Plant operations.
To ensure that the deactivation of systems or components does not result in a reduction of safety, deactivation of systems and components requires the performance of a screening to determine if a 10 CFR 50.59 evaluation is required.
Plant configuration control is implemented through administrative controls. These controls include:
(1) System drawings, as a minimum, should be revised to reflect changes in system lineups and equipment availability.
(2) Procedure revisions are performed as required when systems or components are deactivated.
(3) Removal of components, installation of electrical jumpers, or cutting and capping piping systems will be performed in accordance with Plant design change procedures.
(4) Systems and/or components deactivated should be deenergized and labeled deactivated as appropriate.
4.1-5 Revision 5
Activities applicable to the safe storage of irradiated fuel as recommended in Regulatory Guide Oi 1.33, Revision 2, Appendix A, February 1978, are implemented and maintained by written procedures.
P Oil 1
i l
Revision 5 4.1-6 i
I l
- i. Q SPENT FUFL HANDLING l t
4.2.1 SPENT FUFI RECFIPT. HANDLING. AND TRANSFER
~
l Movement of new or spent fuel into the Reactor Building is not authorized without prior NRC r
3, .
l
- approval.- The movement of spent fuel assemblies within the SFP is the only spent fuel l
. assembly handling activity authorized. This movement would be accomplished by use of the l ;
I 8 : spent fuel pool bridge crane and spent fuel handling tool. Since spent fuel assembly handling l.
is limited to relocating assemblies to alternate storage rack locations, inadvertent criticality l preven'idn is provided as described in Section 4.1.1.
4 4.2.1.1 Functional Descriotion i Figure 4.2-1 provides a layout of the spent fuel storage ein. Movement of spent fuel assemblies is accomplished by use of the spent fuel pool bridge crane and spent fuel handling ,
. tool.
i 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 cf 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 is
- r. manually actuated tool used to hr.ndle 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.
O i 4.2 1 Revision 5 -
. - .._,-.-.,-..o
O 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 from traveling beyond the centerline of the cask load pit. Electrical limit switches deenergize l the bridge drive before the mechanical stops. These stops and limit switches may only bc l bypassed or removed while following an approved procedure, and then only with the Shift l Manager's approval.
4_.2.1.2 Safety Featutt3 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. g Administrative controls in place to minimize the probability and consequences of fuel handhng accidents include:
(1) Fuel handling operations can only be performed under the direct supervision of a CFli.
(2) Fuel handling operations must be performed in accordance with approved Plant procedures.
(3) leads carried over the SFP and the heights at which they may be carried over storage racka containing fuel shall be limited to preclude impact energies over 240,000 in-lbs.
O Revision 4 4.2 2
(4) The Fuel Building crane is restricted to operation within the cask movement envelope shown in Figure 4.2-2 for heavy loads, except when following an <
approved procedure, and then only with the Shift Manager's approval.
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 assemblics, only 781 fuel assemblies are stored. l 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 the operation of the SFP cooling system is provided in Section 4.3.1.
, - Suberitical arrays are ensured by maintaining center to-center distance between adjacent fuel y assemblies and the l'ixed neutron absorber contained in the spent fuel storage racks. Additional discussion is provided in Section 4.1.1.
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 active portion of S fuel assembly. This water barrier serves as a radiation shield enabling the gamma dose rate at the pool surface from the spent assembly to be maintained at or below 2.5 mrem /hr.
p 4.2-3 Revision 5
4.5 RFFERENCES REFERENCES FOR SECTION 4.1 .
- 1. Trojan Nuclear Plant, Final Safety Analysis Report, through Amendment 19 (December 1992).
- 2. PGE 1012. " Trojan Nuclear Plant Fire Protection Program." ;
REFERENCES FOR SECTION 4.2
- 1. Trojan Operating License, No. NPF-1, as amended through Amendment 1% dated l l i
May 19,1997. l l
- 2. Trojan Nuclear Plant, Final Safety Analysis Report, through Amendment 19 (December 1902).
- 3. PGE 1037, " Trojan Nuclear Plant Spent Fuel Storage Rack Replacement Report."
llistorical only. l
- 4. PGE Calculation. Trojan Spent Fuel Pool IIcatup and Boil Down Bases, TC 720 Revision 4, dated November 12, 1996.
REFERENCES FOR SECTION 4.3
- 1. Trojan Nuclear Plant, Final Safety Analysis Report, through Amendment 19 (December 1992).
- 2. PGE 1012. " Trojan Nuclear Plant Fire Protection Program."
Q(3 4.5 1 Revision 5
4 O O O Page 1 of 2 TABLE 4.1-1
! SEISMIC MOMTORING INSTRUMFNTATION e ,
INSTRUMENTS AND SENSOR LOCATIONS MEASUREMENT CIIANNEL CllANWEL CilANWEL NOTES l RANGE CilECK CALIBRATION FUNCT10NAL TEST TRIAXIALTIME-HISTORY RECORDING ACCELEROGRAPHS ST4336A (Containment Base Stab) 0-2 g M 18M SA 1. 3. 4 I ST-6336B '(Containment Wall) 0-2 g M I8M SA 1. 3. 4 ST4336C (Fuel Bldg. Elev.93') 0-2 g M 18M SA 1. 3. 4 ST-6336D (Cable Spreading Roora) 0-2 g M 18M SA 1. 3. 4 ,
ST-6336E (Free Ficid) 0-2 g M iIM SA 1. 3. 4 T2lAXIAL PEAK ACCELEROGRAPHS ;
SR-6340A (Containment Base inside. 45' North) 6-2 g NA 18M NA None SR-6340B (Intake Building Roof. South) 0-2 g NA 18M NA None ,
SR4340C (Top of Containment inside. 201' 9" South) 0-2 g HA 18M NA Nene SR-6340D (Control Building Mezzanine, Top of Ladder 0-2 g NA 18M NA Nonc Above Secondary Sample Storage Room)
SR-6340E (Fuel Building. 93* Ilot Shop. West. Behind the 0-2 g NA 18M NA None Wall) 1' SR-6340F lCCW liest Exchangers.45' Area 3 Base) 0-2 g NA I8M NA None SR-6340G (West EDG Room) 0-2 g NA 18M NA None
, TRIAXIAL RESPONSE-SPECTREIM RECORDERS e SR-634I (Containment Foundation) l NA M 18M SA 2.3.5 Revision 5 i
q
Page 2 of 2 TABLE 4.1-1 SEISMIC MONITORING INSTRUMENTATION I.
NOTES:
- 1. In support of the Triaxial Time-Ilistory Recording Accelerographs, the following instruments shall be operable: SR4336A (Seismic Recorder Unit), SR4336B (Seismic Recorder Unit), SR-6337 (Seismic Playback Unit), and SS-6336 (Seismic Trigger).
- 2. In support of the Triaxial Response-Spectrum Recorders, SAll-6341 (Peak Shock Annunciators) in the Control Room shall be operable.
- 3. Channel check is performed by verifying affected instruments are energized.
- 4. Channel check for Triaxial Time-IIistory Recording Accelerographs does not require a check of the seismic trigger (SS-6336).
- 5. Channel check require verification of Peak Shock Annunciator panel (SAll-6341) indication in Control Room.
The surveillance frequencies are defined as follows:
M = At least once per 31 days SA = At least once per 6 months 18M = At least once per 18 months NA = Not Applicable Revision 5
,O ==>
SPENT FUEL POOL CHEMISTRY SPECIFICATION AND SAMPLING SCHEDULE t
Analysis Value &cQutacy l i Gross Gruuna NA Monthly Gross Ikta NA Quanerly Tritium NA Monthly l J pH 4.0 - 4,7N Weekly l Boron 22000 ppm Weekly
. Conductivity 1 40 us/cm Weekly Chlorida s0.15 ppra Weekly Fluoride :;0.15 ppm Monthly i Calcium r1.0 ppm AR l*l ,
I Magnesium s 1.0 ppm AR 'l '
Suspended Solids s t.0 ppm Monthly
' Sodium s 1.0 ppm Weekly i Demineralizer Effluent
- 4 Gross Gamma NA Monthly
- [a] These analyses would normally be performed as an exploratory measure following other out of specification ,
I conditions, e.g., high sodium or high suspended solids. ;
[b] For Boron levels 22000 ppm, this is an expected wge
- oMy, t-
[NA) Samples obta.aed for trending. No specific limit is applicable.
O Revision 5 i
f
. _ _ _ _ . . _ _ _ . . _ . . _ _ , . . _ _ _ . . - _ ._ . u .. _,_.__.__'
5.2.2 CONTAINMENT VENTil ATION SYSTEM The containment ventilation systems, depicted in Figures 5.2-2 and 5.2 3, provide a means of ventilating the Containment 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 ccntainment purge supply (CS 1) and containment purge exhaust (CS-2).
5.2.2.1 Design B19:3 The design bases for the containment ventilation system are as follows:
(1) Control containment airbonv: 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.
(2) Control and filter the releases of gaseous effluents during containment purge such that the intent of 10 CFR 50, Appendix 1. may be met for overall Plant radioactivity releases.
5.2.2.2 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 ofIIEPA filters.
5.2-3
(4) Two purge supply fans, h (5) Backdraft dampers.
(6) Ductwork, vcives, and instrumentation.
The fans and filters are located outside the containment.
The containment purge exhaust system (CS 2) consists of a the following equipment:
l (1) Prefilters.
(2) A bank of IIEPA filters. l 1
(3) Two purge exhaust fans. h, I \
l 1
(4) Backdraft dampers.
(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.
Revision 5 5.2-4
t i
I t
O
. the limited amount of work anticipated to be performed and the decreased amount of l
~
radioactive material present.
I The HEPA filters installed in the exhaust systems will provide for cleanup of airborne s
radioactive material within the buildings and minimize the release of radioactive particulates to the environment. f t
5.2.4 RADWASTE PROCESSING BUILDING VENTILATION SYSTEM l .
I ,
i The Radwaste Processing Building ventila%n system provides a means of ventilating the l ;
I Radwaste Processing Building and maintaining a negative atmosphere in spaces subject to l
- airborne radioactive contamination. Radwaste storage, handling, and processing operations l f i take place in the building. Portions of the ventilation system have been deactivated. l 5.2.4.1 Desion Bases l
. l The Radwaste Processing Bu!! ding ventilation system is designed to filter exhaust air before l ,
release to the environs, providing high efficiency removal of particulates. Supply air is l provided through infiltration and, as appropriate, through use of the roof supply fans. l l
5.2.4.2 Svetem Descrintion l l
The Radwaste Processing Building ventilation system consists of the following equipment: l 1
J(1) Two e:Anust fans l i
-(2) Supply fans l (3) HEPA filtration l .
~ . :. . (4) Ductwork, instrumentation, and sample taps l
,- l 5.27 Revision 5 i e
!. .,.,,,,,.wJ.-W,-,.. -ey,- ,, r E.,.. , _ . . . _ - , _ , . , , , o---,.. .. ,. . - ., .. m, ___________l
7 O
l The ventilation exhaust ducting contains sanv'c connections that are used with a fixed filter l assembly for collection of air particulates. The sampling arrangement consists of a sample l probe downstream of the llEPA filter, connected through valves and tubing to a filter sampler, l an in line rotameter, and an air pump. An averaged rate of sample flow will be used in the l sample analysis, the frequency of which will be periodically adjusted based on previous l analysis rest . u the type of work currently in progress.
l l Iluilding supply fans are operated as appropriate for workspace habitability. Operation of the l fans is controlled to maintain a negative pressure in the building during radwaste processing l activities.
I l Routine air monitoring is performed during radwaste processing activities in the buildino.
I l Details on sample and analysis schedules are described in the Trojan Nuclear Plant Offsite l Dose Calculation Manual. The schedules comply with the NRC positions described in l Regulatory Guide 1.21.
l l 5.2.4.3 Design Evaluation I
l The Radwaste Processing Building ventilation system provides no safety function. The system l limits the spread or release of airborne radioactive material by filtering the ventilation exhaust l prior to discharge to the environs.
I l The system is operated and maintains a negative atmosphere while the building is occupied.
l The llEPA filter installed in the exhaust duct provides for cleanup of airborne radioactive l material within the building and minimizes the release of radioactive particulates to the l environment.
O )
Revision 5 5.2-8
l l
k The spent resin transfer pump takes a suction from the SRST and discharges back to the SRST and/or to a dispost.ble liner. Normally, the SRST is recirculated to obtain a resin slurry prior to discharge to a disposable liner. A remotely operated resin sampler can be used to obtain a resin sample from the SRST recirculation line.
Water is removed from the disposable liner and returned to the liquid radioactive waste system.
5.4.2.2 Filter llandling in general, potentially contaminated filters will require replacement when filter clogging causes excessive differential pressure across the filter or when radh. tion levels from filter sludges become excessive.
/3 b The filter handling vehicle is provided to allow remote removal and subsequent transfer of highly radioactive expended filters from the filter vessels.
In most cases, 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, Coluinment, l free release facility, or in the Radwaste Processing Building (formerly we Condensate l Demineralizer Building). Some segregation of radwaste material takes place within the plant l buildings. The Radwaste Processing Building is typically used for cutting, decontamination, l and/or packaging solid radioactive waste. l l
c The Radwaste Annex to the Fuel Bu'1 ding includes a drum compactor, which is used for l
/h
() compacting dry, active wastes. A free release facility (formerly the Emergency Diesel l Generator rooms) is used to perform final surveys of decontaminated solid radwaste intended l 5.4-3 Revision 5
l f
l 1 for free release. A solid waste compactor may be used to compact miscellaneous solid waste g materials into drums for storage and shipment offsite.
5.4.3 DESIGN EVALUATION The SRWS does not perform any safety functions. The volume of spent resin required to be processed and stored with the Plant permanently defueled is expected to be less than with the Plant operating. In addition, the maximum expected activity of the spent resin volume is expected to be far less thart 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 filters and miscellaneous solid wastes for the permanently defueled Plant condition are bounded by the analysis for Plant operation. 5 0 N \
Revision 5 5.4-4
1
}
The of0ine gas monitoring systems (PRM 1 and PRM 2) utilize sample pumps to draw the necessary process gas sampics through the sampler. The pumps provide a constant rate of sample Dow inespective of changes in now resistance through the filter media.
Sample lines for offline samplers ase fabricated of stainless steel. Sampling devices and procedures reDect the recommendations of ANSI 13.1 1969, Guide to Sampling Airborne W. An evaluation of sample line losses of Radioactive Material in Nuclear Facilities radiolodine and air particulates was performed for PRM 1 and PRM 2. Line losses were found to be negligible. Each of0ine gas monitoring system is provided with a grab sample connection that can be used to obtain representative samples for laboratory analysis and with fixed filtet assemblies for collection of radiciodine, air particulates, and tritium.
l The following gas monitoring method (s) are also provided:
l (1) Radwaste Processing Building monitoring method.
l l
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 l l
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 partientate channel (PRM 1 A).
f i
5.5 5 Revision 5
PRM 1 utilizes a single sample pump that draws a single gas stream through the air'>orne 9 )
particulate monitoring moving filter. During containment purge operations, a sample is drawn from and returned to the purge exhaust duct by an isokinetic sampling probe. The system may also be aligned to sample the containm:nt atmosphere.
\
5.5.2.2.2 Auxiliary Building Vert Exhaust Monitoring System (PRM-2)
) PRM-2 moi.bors the gaseous and air particu' late activity levels released to the environment by the combiried ventilation exhaust flows from the Fuel and Auxiliary Buildings. The system also mon 5rs releases from the SFP ventilation exhaust and the vent collection header after
= they are diluted by the Fuel and Auxiliary Building ventilation exha.st flow. Section 5.2 3 I= prevides a daription of the Fuel and Auxiliary Building exhaust system.
i
)'
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 tha*. draws a single gas stream in series through the airborne particulate monitoring moving filter and a low- and intermediate-evel gas monitoring chamber. Sample flow is drawn from and teturned to the purge exhaust duct by an isokinetic sampling probe.
- O Revision 5 5.5-6
5.5.2.2.3 SFP Vent Monitoring System (PRM-3)
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 Vent Collection Header Monitoring System (PRM-5)
FRM-5 monitors the gaseous activity levels released to the environment by the verit collection h;ig M '
header exhaust fan. The vent collection header purges various ttnks and sumps containing low level radioactive liquids that could generate airborne radioactivity. Section 5.2.1 provides a description of the vent collection header.
PRM-5 consists of a Nal gamma scintillation detector mounted externally on the process piping. The monitor is positioned downstream of a bigh-efficiency particulate filter to prevent contamination of the process piping at the monitor location by radioactive particulate material.
5.5.2.2.5 It.dwaste Processing Building Monitoring Method l l
This method utilizes a single sample pump that draws a single gas stream through the airborne l l
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. l 5.5-7 Revision 5
M.2.3 Analytical Procedures O)
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.
(4) Liquid scintillation counting.
(5) Radiochemical separations.
I Gross beta analyses are performed with a thin-window proportional counter. Gross alpha analyses are perfonned with a solid-state alpha detector and scaler or thin-window proporticnal counter. Sample volume, counting geometry and counting tirne are chosen to achieve the requisite measurement sensitivities.
Gamma spectrometry is used for isotopic analysis of samples Complex gamma spectra are resolved and an lyzed by computer techniques.
Gaseous tritium concer' rations 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-cxchange columns or by distilling the samples, or both.
Radiochemical procedures are available for the analysis of Sr-89 and Sr-90.
1 Revision 5 5.5-8
l
\,7 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.
5.5.2.4.2 Laboratory Radiation Detectors Counting efficiencies of all laboratory radiation detectors have been detennined 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 eva uated 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 Effluent sampli g 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 (2). The schedules comply with the NRC positions described in Regulatory Guide 1.21W .
O) b 5.5-9 Revision 5 !
l 3.5.4 PROCESS MONITORING AND SAMPLING O
Process sampling and monitoring is used to monitor activity levels within various Plant systems. The sampling freqitency, type of analysis, analytical sensitivity and se purpose of the sample are summarized in Table 5.51 for each liquid process sample location.
ON O}
Revision 5 5.5-10
5.6.3.1.2 Arca Radiation Surveys Radiation protection personnel perform routine rad:ation surveys of all accessible areas of the Plant. These surveys consist of contamination surveys, air samples and external radiation measurements as appropriate for the specific area. Additionally, specific surveys are performed as needed for operational and maintenance functions involving potential exposure of personnel to radiation or radioactive materials.
5.6.3.1.3 Radiation Work Permits All work in radiologically controlled areas is performed under the authorization of radiation work permit 3 issued by radiation protection personnel. These permits state protective clothing l and dosimetry required. monitoring requirements and any special notes or cautions nertinent to the job. These permits also rpecify the maximum levels of contamination, radiation levc!
(including hot spot contact radiation level), and airborne radiation level permitted to be entered under that radiation work permit or will direct the worker to where ruch information is to be obtained.
5.6.3.1.4 Facility Contamination Control Contamination of general Plant areas is controlled by employing various practices and equipment to ensure contamination is controlled at the source to the greatest exts.nt possible.
Additional contaminatior. controls will be specified for jobs involving high levels of contamination (e.g., a double step-off pad. idditional surveys, etc.). Plastic bags and l absorbent paper are used to carty contaminated tools and equipment between r.rers when l required to prevent the spread of contamination. G-M count rate meters (friskers) are located l within the Plant so that personnel can check themselves to determine if they have been contaminated prior to entering another area of the Plant. The final checkpoint for personnel l 9 5.6-7 Revision 3
l leaving the main conaolled area of the Plant is the access control point. Temporary exit points may be established at remote controlled areas as needed.
Airborne antamination is minimized by keeping loose contamina lon levels low and by reducing sources of leakage as much as possible.
Ventilation systems prevent the build-up of air contamination concentrations.
5.6.3.1.5 Personnel Contamination Control 5.6.3.1.5.1 Protective clothing. Contamination of personnel is controlled by the use ot' several types of protective clothing when entering contaminated areas.
l 5.6.3.1.5.2 Respiratory protection. In the event that levels of airborne contamination approach or exceed appiicable limits, provision is made for personnel to use respiratory nrotective equipment. Allowances are made for the use of respiratory protective equipment, as specifically authorized by the USNRC, in determining whether individuals in restricted areas are exposed to concentrations in excess of the values specified in 10 CFR 20.
5.6.3.2 Personnel Dosimetry 5.6.3.2.1 External Radiation Dose Determination TLDs are worn by Plant personnel at all 'imes within radiologically controlled areas to measure radiation dose. All personnel who are assigned TLDs are also required to wear direct-reading pocket ion chambers or a DAD when entering radiologically controlled areas.
The use of DAD readings as a permanent record of an individual's dose will be restricted to O>
Revision 5 5.6-8
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(L1 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT As a result of permarent plant defueling, the production of fission products from prwer-operations has been eliminated. The accident analysis described in this section is based on activity levels associated with defueled conditions. This section describes the bounding accident l scenarios for liquid and gaseous radioactive releases for comparison purposes. Equipment l described in these sections may be deactivated and/or removed. l The Reactor Coolant System (RCS) and secondary system are no longer operated at elevated temperature or pressure. Therefore, releases to the environment from these systems were not considered.
6.1.1 RADIOACTIVE GAS WASTE SYSTEM LEAK OR FAILURE Radioactive waste gas decay tanks (WGDT) permit decay of accumulated radioactive gases prior to their release as s means of reducing the normal release of radioactive materials to the atmosphere. The radioactive contents were principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of halogens.
Since these noble gases are generated from fission during power operation, there will be no generation of fission gases and no more gases sent to the WGDTs from the RCS with the reactor defueled.
The consequences of a WGDT failure were conservatively evaluated assuming gases accumulated prior to permanent plant shutdown had not been released. Reduction in source terms were based solely on decay following permanent Plant shutdown. Resultant offsite doses under these conditions determined that the consequences would be well below 10 CFR 100 limits and EPA PAG limits.
9 6.1-1 Revision f
6.1.1.2 Assumptions on ConditiODS
)
In the evaluation of the WGDT failure accident, the fission product accumulation and release assumptions of Regulatory Guide 1.240 ) were used. The assumptions related to the release of radioactive gases from the postulated failure of a WGDT are:
(1) The reactor had been operating at full power with 1-percent defective fuel and a shutdown to cold condition has been conducted prior to collection of gases in the WGDT.
(2) All noble gases have been removed from the primary cooling system and transferred to the WGDT that is assumed to fail.
(3) The maximum content of the decay tank assumed to fail has been used.
Ol (4) The failure is assumed to occur 40 weeks following Plant shutdown releasing the entire contents of the tank to the Auxiliary Building. The assump:.on of the release of the noble gas inventory from a single tank is based on a design which allows all WGDTs to be isolated from each other when they are in use. No credit was taken for routine tank releases which have occurred in the more than 40 weeks since Plant shutdown.
(5) All of the noble gases are assumed to be exhausted from the Auxiliary Building at ground level over a 2-hour time period. No decay in the Auxiliary Building is assumed.
6.1.1.3 Dose Results Doses resulting from a WGDT failure are conservatively bounded by the Fuel Handling Accident discussed in Section 6.2r2) ,
h 6.1-2
6.3 - SPENT FUEL POOL ACCIDENTS 6.3.1 LOSS OF SPENT FUEL DECAY HEAT REMOVAL CAPABILITY The spent fuel pool (SFP) and the SFP cooling system are designed to: (1) maintain the water in the spent fuel pool at or less than 140'F with the maximum number of fuel assemblies less one full core discharge, (2) maintain fuel cladding integrity in the event all forced cooling is lost and cooling occurs by boiling at the surface of the SFP, with evaporative losses being
! made up by a supply of makeup water, and (3) maintain sufficient cooling of fuel assemblies in the event a fuel assembly or other object is dropped and rests across the top of one or more assembly locations. The only requirement to assure adequate decay heat removal capability for the spent fuel is to maintain the water level in the SFP so that the spent fuel elements remain covered. The design of the SFP is such that a loss of coolant below the top of the fuel is not onsidered to be a credible accident. Events do exist which can result in loss of forced spent fuel cooling or reduce the water inventory in the SFP available for cooling. The worst case
- event allows adequate time (a minimum of ten days) to establish makeup capability to ensure l that fuel elements remain covered.
6.3.1.1 Potential Events Resulting in Lossaf Soent Fuel Decav Heat Removal Capability Chapter 2 identifies those hazards or events which can affect the facility. Section 2.2.3 identifies hazards associated with nearby structures and facilities. These hazards include:
I explosions, toxic chemicals, fires, ship collision with intake stmeture, oil or corrosive liquid spills in the river, and cooling tower collapse.
< Explosive hazards were andyzed for the Trojan site and determined not to result in failure of any safety-related equipment. The SFP is a safety-reb ted structure however the SFP cooling system is not. To be conservative, the explosive hazard was considered to be an initiating
- / event that could result in the damage of the SFP cooling system such N
6.3-1 Revision 5
that a loss of inventory occurs. The consequences of this accident are discussed in ,
Section 6.3.3.
Toxic gas hazards discussed in Section 2.3 consti:ute a hazard to personnel, not Flant equipment. For conservatism, a toxic gas event was assumed to occur with the SFP cooling system out of service. Due to the toxic gas event, personnel are unavailable to restore forced SFP cooling and pool temperature begins to rise. A toxic gas event does not result in a loss of l
SFP cooling system integrity therefore loss of inventory is not considered. The consequences of this accident are discussed in Section 6.3.2.
Fire hazards were also considered. Due to the various support systems required to maintain l forced SFP cooling the Gre is considered to render the SFP cooling system inoperable. SFP cooling system integrity is not considered to be damaged by the Gre therefore loss of inventory is not considered. The consequences of this accident are discussed in S';ction 6.3.2.
O#
The loss of intake structure, either by ship collision or flooding, was evaluated. Forced SFP cooling is considered to be lost by this event due to the loss of the Service Water System (SWS). SFP cooling system integrity is not threatened therefore loss of SFP inventory is not considered. The consequences of this accident are discussed in Section 6.3.2.
Oil or corrosive liquid spills in the river were also considered. As discussed in Section 2.2.3.5, it is unlikely that this event would have any impact on the Trojan Facility.
This event is conservatively bounded by the loss of intake structure discussed above.
Anticipated severe meteorological events have also been evaluated for the Trojan Facility.
These events include tornadoes and high w!nds. Plant structures were designed for the expected meteorological conditions. Section 3.1.3 provides a discussion of the design criteria associated with these meteorological events. For conservatism severe h
6.3-2