ML20195J250
| ML20195J250 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 11/17/1998 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20195J235 | List: |
| References | |
| NUDOCS 9811240221 | |
| Download: ML20195J250 (105) | |
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{{#Wiki_filter:_ ___ ~. 1 TROJAN NUCLEAR PLANT Defueled Safety Analysis Report. Revision 7 t The following information is furnished as a guide for the insertion of new sheets for Revision 7 into the Trojan Nuclear Plant Defueled Safety Analysis Report. This material is denoted by the - use of the revision number in the lower outside corner of the page. Revised pages should be inserted as listed below: Delete (Front /Back) Insert (Front /Back) i/ii through xxix/ blank i/ii through xxix/ blank 1.2-1/1.22 1.2-1/1.2-2 2.2-21/2.2-22 2.2-21/2.2-22 2.3-3/2.3-4 2.3-3/2.3-4 2.3-5/2.3-6 through 2.3-11/2.3-12 2.3-5/2.3-6 through 2.3-9/2.3-10 2.4-31/2.4-32 and 2.4-33/2.4-34 2.4-31/2.4-32 and 2.4-33/2.4-34 2.5-1/2.5-2 2.5-1/2.5-2 3.2-37/3.2-38' 3.2-37/3.2-38 3.3-1/3.3-2 3.3-1/3.3-2 3.3-5/3.3-6 through 3.3-19/ blank 3.3-5/3.3-6 through 3.3-13/3.3-14 Table 3.2-4/3.2-5 Table 3.2-4/3.2-5 Table 3.5-1, Sheets 3/4 and 5/6 Table 3.5-1, Sheets 3/4 and 5/6 Figure 3.2-1/ blank Figure 3.2-1/ blank Figure 3.3-1/ blank Figure 3.3-2/ blank Figure 3.3-3/ blank Figure 3.3-4/ blank 4.0-1/4.0-2 4.0-1/4.0-2 - 4.1-1/4.1-2 and 4.1-3/4.1-4 4.1-1/4.1-2 and 4.1-3/4.1-4 4.3-1/4.3-2 through 4.3-5/4.3-6 4.3-1/4.3-2 through 4.3-5/ blank 4.4-1/ blank 4.4-1/ blank 5.2-9/ blank 5.3-1/5.3-2 through 5.3-7/5.3-8 5.3-1/5.3-1 and 5.3-3/5.3-4 5.4-1/5.4-2 and 5.4-3/5.4-4 5.4-1/5.4-2 and 5.4-3/ blank 5.3-3/5.3-4 5.3-3/5.3-4 5.5-1/5.5-2 through 5.5-5/5.5-6 5.5-1/5.5-2 through 5.5-5/5.5-6 i Table 5.5-1/ blank Table 5.5-1/ blank Figure 5.2-5/5.2-6 Figure 5.2-5/5.2-6 Figure 5.3-1/ blank Figure 5.3-1/ blank Figure 5.3-2/ blank Figure 5.3-3/ blank Figure 5.4-1/ blank 6.3-1/6.3-2 through 6.3-7/ blank 6.3-1/6.3-2 through 6.3-5/6.3-6 g12gga maa,,g N l
L l L. TABLE OF CONTENTS
- (
L DEFUELED SAFETY ANALYSIS REPORT CHAPTER
1.0 INTRODUCTION
AND
SUMMARY
Section Title Page
1.0 INTRODUCTION
AND
SUMMARY
1.0-1 1.1 I nt rod u c t ion..................................... 1.1-1 1.2 General Plant Description............................ 1.2-1 1.2.1 Design Criteria 1.2-1 ) 1.2.2 Fuel Handling System......................... 1.2-2 _1.2.3 Radioactive Waste Treatment Systems.................... 1.2-2 1.3 Identification of Agents and Contractors 1.3-1 ,t \\ 1.3.1 Design of the Facility 1.3-1 1.4 - Exclusions from and Exemotions to Certain Parts of i Title 10 of the Code of Federal Reculations (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 l 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 i 1.4.1.9 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events....... 1.4-3 i Revision 7 l l
CHAPTER
1.0 INTRODUCTION
AND
SUMMARY
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, Appendix H, Reactor Vessel Material Surveillance Program Requirements 1.4-4 1.4.1.12 10 CFR 50.62, Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants.............. 1.4-4 1.4.1.13 10 CFR 50.63, Loss of All Alternating-Current Power....... 1.4-4 1.4.1.14 10 CFR 50.71(e), Maintenance of Record, Making of Reports.. 1.4-5 1.4.1.15 10 CFR 70.24, Criticality Accident Requirements.......... 1.4-5 1.4.2 Exemptions to 10 CFR Related to the Permanently Defueled Condition...................... 1.4-5 1.4.2.1 10 CFR 50.54(o) and Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors 1.4-5 1.4.2.2 10 CFR 50.54(y), Conditions of Licenses................ 1.4-6 1.4.2.3 10 CFR 50.54(q) and Certain Sections of 10 CFR 50.47, " Emergency Plans," 1.4-6 1.5 Material Incornorated by Reference 1.5-1 O Revision 7 ii
CIIAPTER 2.0 SITE CHARACTERISTICS Section Title _P_ age _ 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 Map................................. 2.1-2 2.1.1.3 Boundaries for Establishing Effluent Release Limits......... 2.1-3 2.1.2 Exclusion Area Authority and Control.................... 2.1-3 2.1.2.1 ~ Au tho rity.................................... 2.1-3 2.1.2.2 Exclusion Area Activities Unrelated to Plant Operation....... 2.1-4 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 10 Miles......................... 2.1-6 2.1.3.2_ Population Between 10 and 50 Miles................... 2.1-8 2.1.3.3 Transient Populatic. 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. Transportation and and Military Facilities..... 2.2-1 2.2.1 Locations and Routes............................... 2.2-1 2.2.2 Descriptions 2.2-4 2.2.2.1 Description of Products and Materials.................. 2.2-4 2.2.2.2 Pipe li ne s..................................... 2.2-5 2.2.2.3 Waterways 2.2-6 l 2.2.2.4 A i rpo rt s..................................... 2.2-6 2.2.3 Evaluation of Potential Accidents....................... 2.2-7 2.2.3.1 Explosions 2.2-7 2.2.3.2 Toxic Chemicals................................ 2.2-17 -2.2.3.3 Fires 2.2-19 2.2.3.4 Ship Collision with Intake Structure 2.2-20 p 2.2.3.5 Oil or Corrosive Liquid Spills in River..... 2.2-21 J iii Revision 7 r
1 CHAPTER 2.0 SITE CHARACTERISTICS Section Title Page 2.2-22 2.2.3.6 Cooling Tower Collapse........................... 2.3-1 2.3 Meteorology.................................... 2.3-1 2.3.1 Regional Climatology.............................. 2.3.1.1 General Climate................................ 2.3-1 2.3.1.2 Regional Meteorological Conditions for Design and Operation Bases...................... 2.3-1 2.3-2 2.3.2 Local Meteorology................................ 2.3.2.1 Normal and Extreme Values of Meteorological Parameters 2.3-2 2.3.2.2 PotentialInfluence of the Plant and its Facilities On Local Meteorology................ 2.3-4 2.3.2.3 Local meteorological Conditions for Design and Operation Bases....................... 2.3-5 2.3.3 Onsite Meteorological Measurements Program.............. 2.3-5 l 2.3.4 Diffusion Estimates................................ 2.3-6 2.4 Hydrologic Engineering............................. 2,4-1 2.4-2 2.4.1 Hydrologic Description...... 2.4.1.1 Site and Facilities............................... 2.4-2 2.4.1.~2 Hydrosphe re.................................. 2.4-2 2.4-3 2.4.2 Floods 2.4-3 2.4.2.1 Flood History 2.4.2.2 Flood Design Considerations..........,............. 2.4-4 2.4.2.3 Effects of LocalIntense Precipitation 2.4-5 2.4.3 Probable Maximum Flood of Streams and Rivers............. 2.4-5 2.4.3.1 Probable Maximum Precipitation........... 2.4-5 2.4.3.2 Precipitation Losses.............................. 2.4-8 2.4.3.3 Runoff Model 2.4-9 2.4.3.4 Probable Maximum Flood Flow................. 2.4-12 l Revision 7 iv i j
p CHAPTER 2.0 Q SITE CHARACTERISTICS Section Title Page 2.4.3.5 Water Level Determinations........................ 2.4-18 2.4.3.6 Coincident Wind Wave Activity...................... 2.4-19 2.4.4 Potential Dam Failures.............................. 2.4-20 2.4.4.1 Seismically Induced Dam Failure..................... 2.4-20 I 2.4.4.2 Volcanically Induced Dam Failure.................... 2.4-22 2.4.4.3 Spirit Lake Blockage Failure........................ 2.4-25 2.4.5 Probable Maximum Surge Flooding................... 2.4-29 2.4.5.1 Surge Water Levels.............................. 2.4-29 2.4.5.2 Resonance.................................... 2.4-30 2.4.6 Probable Maximum Tsunami Flooding................... 2.4-31 2.4.7 Ic e Effec ts...................................... 2.4-31 2.4.8 Cooling Water Canals and Reservoirs.................... 2.4-32 2.4.9 C ha nne l D ive rs io ns................................ 2.4-32 2.4.10 Flooding Protection Requirements...................... 2.4-32 2.4.11 Low Water Considerations........................... 2.4-33 2.4.12 Dispersion, Dilution, and Travel Times of Accidental Releases of Liquid Effluents in Surface Waters............. 2.4-33 2.4.13 G rou nd wa te r.................................... 2.4-34 2.4.13.1 Description and Onsite Use.................. 2.4-34 2.4.13.2 S ou rc es...................................... 2.4-35 2.4.13.3 Accide nt Effec ts................................ 2.4-36 2.5 Geology. Seismology and Geotechnical Engineering........... 2.5-1 2.5.1 Basic Geologic and Seismic Information 2.5-1 2.5.1.1 Regional Geology............................... 2.5-2 2.5.1.2 Site Geology.................................. 2.5-4 f y Revision 7
CHAPTER 2.0 SITE CHARACTERISTICS Section Title Page 2.5.2 Vibratory Ground Motion............................ 2.5-8 2.5.2.1 Se ismicity.................................... 2.5-8 2.5.2.2 Geologic Structures and Tectonic Activity............... 2.5-8 2.5.2.3 Maximum Earthquake Potential...................... 2.5-9 2.5.2.4 Seismic Margin Earthquake, Safe Shutdown Earthquake and Operating Basis Earthquake..................... 2.5-11 2.5.3 Surface Faulting.................................. 2.5-13 2.5.3.1 Geologic Condition of the Site....................... 2.5-13 2.5.3.2 Investigation for Capable Faults...................... 2.5-14 2.5.3.3 Description of Faults............................. 2.5-16 l 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 2.5-33 2.5.6 Volcanology 2.5.6.1 Volcanoes in General Area......................... 2.5-34 2.5.6.2 Possible Hazards of the Cascade Volcanoes 2.5-38 2.5.6.3 Hazards from Future Volcanic Activities................ 2.5-39 2.5.6.4 Summary and Conclusions 2.5-47 2.6 References 2.6-1 Revision 7 vi
r-CHAPTER 3.0 ( FACILITY DESIGN Section Title Page 3.0 FACILITY DESIGN............ 3.0-1 3.1 Summuy...................................... 3.1-1 3.1.1 Conformance with NRC Ge.neral Design Criteria............. 3.1-1 3.1.2 Classification of Structures, Components and Systems 3.1-7 3.1.3 Wind and Tornado Loadings......................... 3.1-7 3.1.3.1 Wind Load ings................................. 3.1-8 3.1.3.2 Tornade Leadings............................... 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 Missile Selection and Description..................... 3.1-14 AV 3.1.6 Seismic Design................................... 3.1-15 1 l 3.1.6.1 Seismic Design Parameters......................... 3.1-15 l 3.1.6.2 Seismic System Analysis 3.1-19 3.2 Spent Fuel Storage 3.2-1 3.2.1 Control, Auxiliary, and Fuel Building Complex 3.2-1 3.2.1.1 Descrip tion................................... 3.2-1 3.2.1.2 Des ig n Bases.................................. 3.2-6 3.2.1.3 Applicable Codes, Standards, and Specifications........... 3.2-8 3.2.1.4 Loads and Load Combinations....................... 3.2-13 3.2.1.5 Design and Analysis Procedures...................... 3.2-19 l 3.2.1.6 Structural Acceptance Criteria....................... 3.2-28 3.2.1.7 Materials, Quality Control and Special Construction Techniques........ 3.2-30 3.2.1.8 Testing and Inservice Inspection Paquirements............ 3.2-34 l 3.2.1.9 Foundations................................... 3.2-34 3.2.2 Spent Fuel Pool and Fuel Storage Racks 3.2-35 3.2.2.1 Design Bases............... 3.2-35 ( vii Revision 7 f il-
CHAPTER 3.0 FACILITY DESIGN Section Title Page 3.2.2.2 System Desig 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-evaluation for Beyond Design Basis Seismic Motions...................... 3.2-40 3.3 Auxiliarv s ystems................................. 3.3-1 3.3.1 Fuel Handling System........................ 3.3-1 3.3.1.1 Des ign Bases.................................. 3.3-1 3.3.1.2 System Description.............................. 3.3-2 3.3.1.3 Design Evaluation............................... 3.3-4 3.3.2 Modular SFP Cooling and Cleanup System................. 3.3-5 3.3.2.1 Design Bases......................... 3.3-5 3.3.2.2 System Description.............................. 3.3-6 h 3.3.2.3 Design Evaluation............................... 3.3-8 l 3.3.3 Deleted 3.3.4 Service Water System.............................. 3.3-9 3.3.4.1 Design Bases...... 3.3-9 3.3.4.2 System Description..................... 3.3-10 I 3.3.5 Compressed Air System............................. 3.3-11 3.3.6 Boric Acid Batch Tank............. 3.3-12 3.3.7 Makeup Water Treatment System..... 3.3-12 3.3.8 Equipment and Floor Drain Systems.......... 3.3-12 3.3.9 Plant Discharge and Dilution Structure 3.3-13 1 l 3.3.10 Deleted Revision 7 viii
_= ___. - -.- l lb CHAPTER 3.0 'a FACILITY DESIGN Section Title Page 3.3.11 Fire Protection System and Program..................... 3.3-14 3.3.12 Control Room Habitability........................... 3.3-14 1 t 3.3.13 Seismic Instrumentation............................. 3.3-14 ) 3.4 Electric Power................................... 3.4-1 3.4.1 Offsite Power System 3.4-1 i 3.4.1.1 Descrip tion................................... 3.4-1 3.4.1.2 Ana lys is..................................... -3.4-2 3.4.2 Onsite Power Systems.............................. 3.4-2 l 3.4.2.1 Description.................................... 3.4-2 3.4.2.2 Analy s i s..................................... 3.4-4 l 3.5 Compliance with NRC Regulatory Guides 3.5-1 3.6 References 3.6-1 l l f-l l l l i 4 '=( ix Revision 7 i f t
CHAPTER 4.0 OPERATIONS O Section Title Page 4.0 OPERATI ON S................................... 4.0-1 4.1 Ooeration Descriotion.............................. 4.1-1 4.1.1 Criticality Prevention............................... 4.1-1 4.1-2 4.1.2 Chemistry Control 4.1.3 Itstrumentation 4.1-3 4.1.3.1 , Seismic Monitoring Instrumentation................... 4.1-3 4.1.4 Maintenance Activities..................... 4.14 4.1.5 Administrative Control of Systems...................... 4.1-5 4.2 Spent Fuel Handling............................... 4.2-1 4.2.1 Spent Fuel Receipt, Handling, and Transfer................ 4.2-1 4.2.1.1 Functional Description............................ 4.2-1 4.2.1.2 Safety Features................................. 4.2-2 4.2.2 Spent Fuel Storage................................ 4.2-3 4.3 Spent Fuel Cooling and Support Systems.................. 4.3-1 4.3.1 Spent Fuel Pool Cooling............................ 4.3-1 4.3.1.1 Off-Normal Operation of the Spent Fuel Cooling System...... 4.3-2 4.3.1.2 Loss of Spent Fuel Pool Level....................... 4.3-2 4.3.1.3 Loss of Spent Fuel Pool Cooling 4.3-3 4.3.1.4 High Spent Fuel Pool Level 4.3-4 4.3.1.5 Safety Criteria and Assurance....................... 4.3-4 4.3.2 Electrical Distribution.... 4.3-5 1 4.4 Control Room Area........................ 4.4-1 4.5 References 4.5-1 Revision 7 x
L I I-CHAPTER 5.0 i RADIATION PROTECTION' t i ' Section Title Page 5.0 RADIATION PROTECTION......................... 5.0-1 l 5.1 Source Terms 5.1-1 5.2 Offgas Treatment and Ventilation....................... 5.2-1 5.2.1 Vent Collection System............................. 5.2-1 l 5.2.1.1 . Design Bases.................................. 5.2-1 l 5.2.1.2 System Description.............................. 5.2-2 5.2.1.3 - Design Evaluation............................... 5.2-2 5.2.2 Containment Ventilation System........................ 5.2-3 . 5.2.2.1 Des ig n Bases.................................. 5.2-3 5.2.2.2 System Description.............................. 5.2-3 5.2.2.3 Design Evaluation............. s 5.2-5 5.2.3 Fuel and Auxiliary Building Ventilation System.............. 5.2-5 l 5.2.3.1 Design Bases.................................. 5.2-5 l p) 5.2.3.2 System Description.............................. 5.2-5 [ 5.2.3.3 Design Evaluation............................... 5.2-6 o 5.2.4 Radwaste Processing Building Ventilation System............ 5.2-7 i - 5.2.4.1 Des ig n Bases.................................. 5.2-7 5.2.4.2 System Description.............................. 5.2-7 5.2.4.3 Design Evaluation............................... 5.2-8 5.2.5 Modular Spent Fuel Pool Cooling System Cooling Air......... 5.2-9 5.3 Liauid Waste Treatment and Retention 5.3-1 5.3.1 De s ig n B as es.................................... 5.3-1 5.3.2 System Description................................ 5.3-2 - 5.3.3 Design Evaluation................................. 5.3-4 5.4 Sol id Wa s te s.................................... 5.4-1 i ~ xi Revision 7 l i. l l v
1 CHAPTER 5.0 hi RADIATION PROTECTION Section Title Page 5.4.1 Design B ases.................................... 5.4-1 5.4.2 System Description................................ 5.4-2 5.4.2.1 Spent Resin Transfer System........................ 5.4-2 5.4.2.2 Filter Handling................................. 5.4-2 5.4.2.3 Solid Was tes.................................. 5.4-3 5.4.3 Design Evaluation................................. 5.4-3 5.5 Process and Effluent Monitoring Systems.................. 5.5-1 5.5.1 D es ig n B as es.................................... 5.5-1 5.5.2 System Description.................. 5.5-2 5.5.2.1 Liquid Monitoring Systems..................... 5.5-3 5.5.2.2 Gas Monitoring Systems........................... 5.5-4 5.5.2.3 Analytical Procedures............................ 5.5-8 5.5.2.4 Calibration and Maintenance........................ 5.5-9 5.5.3 Effluent Monitoring and Sampling...................... 5.5-9 5.5.4 Process Monitoring and Sampling 5.5-10 5.6 Radiation Protection Program......................... 5.6-1 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 Monitoring Instrumentation 5.6-1 5.6.2 Equipment. Instrumentation and Facilities 5.6-2 5.6.2.1 Radiation Protection Facilities...................... 5.6-3 5.6.2.2 Radiation Protection Instrumentation.............. 5.6-4 5.6.3 Procedures... 5.6-6 5.6.3.1 Control of Personnel Radiation Exposure.... 5.6-6 5.6.3.2 Personnel Dosimetry... 5.6-8 Revision 7 xii
I l i CHAPTER 5.0 i RADIATION PROTECTION ' g ( Section Title Page 5.6.3.3 Radioactive Materials Safety........................ 5.6-9 5.7 - Ensuring that Occupational Radiation Exnosures are a low as is Reasonably Achievable (ALARA) 5.7-1 5.7.1 Policy Considerations.............................. 5.7-1 l 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 I l I-I' r 4 xiii Revision 7 i
CHAPTER 6.0 ACCIDENT ANALYSIS Section Title Page 6.0 Accident Analysis................................. 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 Assump tions.................................. 6.0-5 4 6.0.2.2 Whole Body Dose............................... 6.0-5 6.0.2.3 Thyroid Inhalation Dose........................... 6.0-7 6.0.2.4 Computer Code................................ 6.0-8 6.1 Radioactive Release from a Subsystem or Component.......... 6.1-1 6.1.1 Radioactive Gas Waste System Leak or Failure.. 6.1-1 6.1.1.2 Assumptions or Conditions......................... 6.1-2 6.1.1.3 Dose Results.................................. 6.1-2 6.1.2 Postulated Radioactive Releases Due to Liquid Tank Failures................... 6.1-3 6.1.2.1 Identification of Causes and Accident Description.......... 6.1-3 6.1.2.2 Do se Results.................................. 6.1-4 6.2 Fuel Handling Accident 6.2-1 6.2.1 Assumptions or Conditions........ 6.2-1 6.2.2 D o s e Res ults.................................... 6.2-3 6.3 Spent Fuel Pool Accidents 6.3-1 6.3.1 Loss of Spent Fuel Decay Heat Removal Capability........... 6.3-1 6.3.1.1 Potential Events Resulting in Loss of Spent Fuel Decay Heat Removal Capability 6.3-1 6.3.2 Loss of Forced Spent Fuel Cooling Without Concurrent SFP Inventory Loss.............. 6.3-4 Revision 7 xiv
I i l CHAPTER 6.0 ( ACCIDENT ANALYSIS l Section Title Page l l 6.3.3 Loss of Forced Spent Fuel Cooling with Concurrent l SFP Inventory Loss.............................. 6.3-5 l 6.4 - References 6.4-1 l l i t 8 l l I l ? i i i l.- i t f I \\ xv Revision 7
CHAPTER 7.0 CONDUCT OF OPERATIONS Section Title Page 7.0 CONDUCT OF OPERATIONS........................ 7.0-1 7.1-1 7.1 PGE Organizational Structure......................... 7.1.1 Management and Technical Support Organization 7.1-1 7.1-1 7.1.2 Nuclear Divisian................................. 7.1-1 7.1.2.1 Plant Organizations.............................. 7.1-3 7.1.2.2 Supporting Organizations.......................... 7.1.2.3 Review and Audit Organizations 7.1-5 7.2 P l a n t Proc ed u re s.................................. 7.2-1 7.2.1 Procedures Related to Nuclear Safety.. 7.2-1 7.2.2 Nuclear Division Manual............................ 7.2-2 7.2.3 Plant Procedures....... 7.2-3 7.2.3.1 Administrative Orders.... 7.2-4 7.2-4 7.2.3.2 Operating Instructions........................... 7.2.3.3 Periodic Tests 7.2-4 7.2-4 7.2.3.4 Fuel Handling Procedures 7.2.3.5 Maintenance Procedures........................... 7 7-5 7.2.3.6 Radiation Protection Procedures......... M 7.2.3.7 Chemistry Procedures............................ 7.2-5 7.2.3.8 Plant Safety Procedures........ 7.2-6 7.2.3.9 Temporary Procedures............................ 7.2-6 7.3-1 7.3 Training 7.3.1 Training for Certified Fuel Handlers.................. 7.3-1 7.3.2 Training for Plant Staff............. 7.3-1 7.3.2.1 General Employee Training....... 7.3-1 7.3.2.2 Fire Brigade Training 7.3-2 7.3.2.3 Other Training Programs... 7.3-2 7.4 Emergencv Plan............ 7.4-1 Revision 7 xvi
1 l CHAPTER 7.0 CONDUCT OF OPERATIONS \\ Section Title Page 7.5 Decommissioning Plan.............................. 7.5-1 7.6 Troian Nuclear Plant Security Plan /Troian Nuclear Plant Security Force Training and Oualification Plan.................... 7.6-1 n i xvii Revision 7 l
CHAPTER 8.0 g TECHNICAL SPECIFICATIONS Section Title Page 8.0 TECHNICAL SPECIFICATIONS...................... 8.0-1 l i f 9 Revision 7 xviii
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-M-r4Jk 4 M 4--'---&*----- = - - * -+h M L-4- J 8M 'd- '*M4* CHAPTER 9.0 QUALITY ASSURANCE Section Title Page 9.0 QUALITY ASSURANCE........................... 9.o.1 d t e r i i 4 a 1 4 a J i n 4 xix Revision 7 4
LIST OF TABLES l DEFUELED 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) 2.3-4 Annual Average x /Q, Deposition and Plume Depletion Factor at Site Boundaries and Offsite Exposure Locations for Ground Level Releases (Trojan Site Data September 1,1972 - August 31,1974) 2.3-5 Maximum Annular Sector, Terrain Adjustment Factors Derived From NUSPUF With Buildir.g Wake Adjustment Divided by NUSOUT for Standard Population Distances of 0.5 Mile to 4.5 Miles 2.3-6 Terrain Adjustment Factors Derived form NUSPUF With Building Wake Adjustment Divided by NUSOUT for Special Instances 3.1-1 Design Wind Loads t l 3.1-2 Capability of Structures to Withstand a Tornado g 3.1-3 Locations of Gas Storage Tanks W l 3.1-4 Analyzed SFP Load Drops and Missiles l 3.1-5 Damping Values, Percent Critical Damping l 3.1-6 Frequencies and Modal Effective Weights for Control, Auxiliary, and Fuel l Building Complex 3.1-7 Comparison of Maximum Accelerations from Time-History and Response Spectrum Analyses of the Unmodified Control Building Complex 3.2-1 Live Loads 3.2-2 Calculated Results - Fuel Building 3.2-3 Calculated Results - Auxiliary Building 3.2-4 Calculated Results - Control Building 3.2-5 Significant Elevations in the Spent Fuel Pool 3.5-1 List of Pertinent Regulatory Guides for Defueled Status 4.3-1 Spent Fuel Pool Chemistry Specification and Sampling Schedule i i 5.5-1 Radiological Analysis Summary of Liquid Process Samples Revision 7 xx
LIST OF TABLES DEFUELED SAFETY ANALYSIS REPORT Number Title 6.0-1 Core and Gap Activities Based on Full Power Operation for 650 Days Full Power: 3565 MWt 6.0-2 Core Temperature Distribution 6.0-3 Breathing Rates 6.0-4 Physical Data for Isotopes 6.0-5 Accident Atmospheric Dilution Factors 6.2-1 Input Data for Calculation of Site Boundary Doses of a Fuel Handling Accident 6.2-2 Resultant Doses from Fuel Handling Accident and Comparison with 10 CFR 100 6.3-1 Spent Fuel Pool Performance During Loss of Forced Cooling 6.3-2 Dose Rates at Spent Fuel Pool at Reduced Water Levels G,V l V xxi Revision 7
LIST OF FIGURES DEFUELED SAFETY ANALYSIS REPORT Number Title 1.1-1 Facility Location 2.3-1 Trojan Accident Meteorology 2.4-1 Lower Columbia River Probable Maximum Flood, Comparative Hydrographs for Columbia River at the Dalles 2.4-2 Lower Columbia River Probable Maximum Flood Source of Flood Volumes of Major Floods 3.1-1 Design Response Spectra Operating Basis Earthquake 3.1-2 Design Response Spectra Safe Shutdown Earthquake 3.1-3 Synthetic Acceleration Time History Normalized to 1.0 g 3.1-4 Comparison of the Acceleration Response Spectrum of the Synthetic Time History With the Trojan Design Spectrum for 1% Damping 3.1-5 Comparison of the Acceleration Response Spectrum of the Synthetic Time History With the Trojan Design Spectrum for 2% Damping l 3.1-6 Comparison of the Acceleration Response Spectrum of the Synthetic Time History With the Trojan Design Spectrum for 5% Damping g 3.1-7 Comparison of the Acceleration Response Spectrum of the Synthetic Time W History With the Trojan Design Spectrum for 7% Damping 3.1-8 Finite Element Model - Control, Auxiliary, and Fuel Building 3.1-9 North - South Horizontal Response Spectra for Control Building Elevations 61 feet 0 inch and 65 feet 0 inch 3.1-10 North - South Horizontal Response Spectra for Auxiliary Building ( Elevation 61 feet 0 inch 3.1-11 North - South Horizontal Response Spectra for Fuel Building Elevation 61 feet 0 inch 3.1-12 North - South Horizontal Response Spectra for Control Building Elevation 117 feet 0 inch 3.1-13 North - South Horizontal Response Spectra for Auxiliary Building l Elevation 117 feet 0 inch l 3.1-14 East - West Horizontal Response Spectra for Control Building Elevations l 61 feet 0 inch and 65 feet 0 inch 3.1-15 East - West Horizontal Response Spectra for Auxiliary Building Elevation 61 feet 0 inch 3.1-16 East - West Horizontal Response Spectra for Fuel Building Elevations 61 feet 0 inch Revision 7 xxii
LIST OF FIGURES DEFUELED SAFETY ANALYSIS REPORT - Number Title 3.1-17 East - West Horizontal Response Spectra for Control Building Elevation 117 feet 0 inch 3.1-18 East - West IIorizontal Response Spectra for Auxiliary Building Elevation 117 feet 0 inch 3.1-19 3-D Model for the Fuel Building Steel Superstructure 3.2-1 Fuel Building - Plan Elevation 45' 3.2-2 Spent Fuel Pool - Typical Details 3.2-3 Auxiliary Building - Plan Elevation 45' 3.2-4 Auxiliary Building - Plan Elevation 45', Containment Abutment 3.2-5 Fuel Building - Plan Elevation 66' 3.2-6 Auxiliary Building - Plan Elevation 61' 3.2-7 Typical Section Through Auxiliary and Fuel Buildings J 3.2-8 Typical Steel Framing 3.2-9 Typical Steel Column Details 3.2-10 Floor Plans Showing Modifications 3.2-11 Control Building Floor Plan EL 45'-0" Showing Existing and Shear Walls ,( 3.2-12 Control Building Floor Plan EL 61'-0" & 65'-0" Showing Existing and Shear Walls 3.2-13 Control Building Floor Plan EL 77'-0" Showing Existing and Shear _ Walls 3.2-14 Control Building Floor Plan EL 93'-0" Showing Existing and Shear Walls 3.2-15 Equipment Location, Reactor and Auxiliary Buildings - Plan Below Ground Floor 4 3.2-16 Equipment Location, Reactor and Auxiliary Buildings - Plan Operating Floor, Elevation 45' 3.2-17 Equipment Location, Reactor and Auxiliary Buildings - Plan Elevation 61' 3.2-18 Equipment Location, Reactor and Auxiliary Buildings - Plan Elevation 77' 3.2-19 Equipment Location, Reactor and Auxiliary Buildings - Plan Operating Floor and Above 3.2-20 Equipment Location, Reactor and Auxiliary Buildings - Section A-A 3.2-21 Equipment Location, Reactor and Auxiliary Buildings - Sections B-B, D-D, E-E, F-F 3.2-22 Equipment Location, Reactor and Auxiliary Buildings - Sections C-C and F-F 3.2-23 Containment Structure Typical Details 3.2-24 Containment Structure Typical Liner Plate Details 3.2-25 Containment Structure Base Slab Bottom Reinforcing xxiii Revision 7 l J
1 LIST OF FIGURES DEFUELED SAFETY ANALYSIS REPORT Number Title 3.2-26 Containment Structure Base Slab Top Reinforcing 3.2-27 Containment Structure Wall Reinforcing 3.2-28 Containment Structure Dome 3.2-29 Containment Structure Typical Penetration Details 3.2-30 Containment Structure Prestressing Tendons at Equipment Hatch I 3.3-4 Modular Spent Fuel Pool Cooling and Cleanup System 4.2-1 Spent Fuel Storage Pool 4.2-2 Plant Arrangement Diagram of Fuel Cask Movement Envelope 5.2-1 Gaseous Radioactive Waste System 5.2-2 Containment Purge Supply System (CS-1) 5.2-3 Containment Purge Exhaust System (CS-2) 5.2-4 Fuel / Auxiliary Building Ventilation Supply System (AB-2) 5.2-5 Fuel / Auxiliary Building Ventilation Exhaust System (AB-3) 5.2-6 SFP Ventilation Exhaust System (AB-4) g 5.2-7 Condensate Demineralizer Building Ventilation Exhaust System W 5.3-1 Liquid Radioactive Waste System I 6.3-1 Decay Heat Generated from Stored Fuel 6.3-2 SFP Heatup Rate versus Time After Reactor Shutdown 6.3-3 Time for SFP to Boil Upon IAss of Forced Cooling 6.3-4 SFP Boil Off Rate Without Makeup versus Time After Reactor Shutdown 6.3-5 Makeup Rate to Maintain SFP Level During Boil Off versus Time After Reactor Shutdown 6.3-6 Boil Off Time to 10 Feet Above Fuel Versus Time After Reactor Shutdown Revision 7 xxiv
l LIST OF EFFECTIVE PAGES .p V DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision - Title Page N/A Rev.0 Table of Contents i - xxix Rev.7 1.0 1.0-1 Rev.0 1.1 1.1-1 Rev.0 1.2 1.2-1 Rev.0 1.2 1.2-2 Rev.7-1.3 1.3-1 Rev.0 1.4 1.4-1 through 1.4-4 Rev.4 1.4 1.4-5 Rev.6 1.4 1.4-6 Rev.4 1.5 1.5-1 Rev.5 Figure 1.1-1 N/A Rev.0 ' 2.0 2.0 Rev.0 l 2.1 2.1-1 and 2.1-2 Rev.O O 2.1 2.1-3 Rev.3 V 2.1 2.1-4 through 2.1-12 Rev.0 2.2 2.2-1 through 2.2-7 Rev.O l 2.2 2.2-8 Rev.4 2.2 2.2-9 Rev.0 2.2 2.2-10 Rev.4 2.2 2.2-11 through 2.2-16 Rev.0 2.2 2.2-17 Rev.3 2.2 2.2-18 and 2.2-19 Rev.0 2.2 2.2 20 Rev.6 2.2 2.2-21 Rev.7 2.2 2.2-22 Rev.0 2.3 2.3-1 and 2.3-2 Rev.O { 2.3 - 2.3-3 Rev.7 1 2.3 2.3-4 Rev.O l 2.3 2.3-5 through 2.3-10 Rev.7 { 2.4 2.4-1 Rev.3 l 2.4 2.4-2 and 2.4-3 Rev.0 2.4 2.4-4 Rev.3 1 2.4 2.4-5 through 2 4-22 Rev.0 2.4 2.4-23 Rev.6 I 't xxv Revision 7 i
LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision 2.4 2.4-24 through 2.4-31 Rev.0 2.4 2.4-32 and 2.4-33 Rev.7 2.4 2.4-34 and 2.4-35 Rev.0 2.4 2.4-36 Rev.6 2.5 2.5-1 Rev.7 2.5 2.5-2 through 2.5-8 Rev.0 2.5 2.5-9 Rev.6 2.5 2.5-10 and 2.5-11 Rev.0 2.5 2.5-12 Rev.6 2.5 2.5-13 Rev.4 2.5 2.5-14 through 2.5-47 Rev.0 2.6 2.6-1 through 2.6-10 Rev. O Tables 2.3-1 through 2.3-6 N/A Rev.O Figure 2.3-1 N/A Rev.0 Figures 2.4-1 and 2.4-2 N/A Rev.0 3.0 3.0-1 Rev.0 3.1 3.1-1 through 3.1-5 Rev.0 3.1 3.1-6 and 3.1-7 Rev.5 3.1 3.1-8 through 3.1-28 Rev.0 3.2 3.2-1 Rev.0 3.2 3.2-2 Rev.4 3.2 3.2-3 through 3.2-7 Rev.5 l 3.2 3.2-8 through 3.2-34 Rev.4 3.2 3.2-35 through 3.2-37 Rev.5 3.2 3.2-38 Rev.7 3.2 3.2-39 through 3.2-41 Rev.5 3.3 3.3-1 Rev.7 3.3 3.3-2 through 3.3-3 Rev.0 3.3 3.3-4 Rev.4 3.3 3.3-5 through 3.3-14 Rev.7 3.4 3.4-1 Rev.0 3.4 3.4-2 and 3.4-3 Rev.4 3.4 3.4-4 Rev.6 3.5 3.5-1 Rev.0 3.6 3.6-1 Rev.0 3.6 3.6-2 Rev.5 Revision 7 xxvi
_. _ _.. ~ _. _ _ _ _ _ _ _.. !O LIST OF EFFECTIVE PAGES i f-DEFUELED SAFETY ANALYSIS REPORT t Section Effective Pages Revision I-3.6 3.6-3 Rev.O L Tables 3.1-1 and 3.1-2 N/A Rev.O Table 3.1-3 N/A Rev.6 Tables 3.1-4 through 3.1-7 N/A Rev.O Tables 3.2-1 through 3.2-3 N/A Rev.0 Table 3.2-4 N/A Rev.1 Table 3.2-5 N/A Rev.7 Table 3.5-1 Sheet 1 Rev.3 Table 3.5-1 Sheet 2 Rev.4 Table 3.5-1 Sheets 3 through 5 Rev.7 Table 3.5-1 Sheets 6 through 8 Rev.3 Table 3.5-1 Sheets 9 through 15 Rev.4 Table 3.5-1 Sheets 16 through 38 Rev.3 Figures 3.1-1 through 3.1-19 N/A Rev.0 Figures 3.2-1 N/A Rev.7 Figures 3.2-2 through 3.2-13 N/A Rev.0 - Figure 3.2-14 N/A Rev.6 Figures 3.2-15 through 3.2-22 N/A Rev.4 Figures 3.2-23 through 3.2-30 N/A Rev.0 Figure 3.3-4 N/A Rev.7 4.0 4.0-1 and 4.0-2 Rev.7 4.1 4.1-1 through 4.1-4 Rev.7 4.1 4.1-5 and 4.1-6 Rev.6 ' 4.2 4.2-1 Rev.5 4.2 4.2-2 Rev.4 - 4.2 4.2-3 Rev.5 4.3 4.3-1 through 4.3-3 Rev 7 4.3 4.3-4 Rev.6 4.3 4.3-5 Rev.7 4.4 4.4-1 Rev.7 4.5 4.5-1 Rev.5 Table 4.3-1 N/A-Rev.5 Figure 4.2-1 N/A Rev.4 l. Figure 4.2-2 N/A Rev.4 L 5.0 5.0-1 Rev.0 l \\ xxvii Revision 7 L e .m..
4 LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORI Section Effective Pages Revision 5.1 5.1-1 Rev.0 5.2 5.2-1 through 5.2-3 Rev.0 5.2 5.2-4 Rev.5 5.2 5.2-5 and 5.2-6 Rev.0 5.2 5.2-7 and 5.2-8 Rev.5 5.2 5.2-9 Rev.7 5.3 5.3-1 through 5.3-4 Rev.7 5.4 5.4-1 through 5.4-3 Rev.7 5.5 5.5-1 Rev.0 5.5 5.5-2 through 5.5-6 Rev.7 5.5 5.5-7 through 5.5-10 Rev.5 5.6 5.5-1 Rev.1 5.6 5.6-2 Rev.3 5.6 5.6-3 Rev.0 5.6 5.6-4 through 5.6-7 Rev.3 5.6 5.6-8 Rev.5 5.6 5.6-9 and 5.6-10 Rev.3 g 5.7 5.7-1 Rev.O 5.8 5.8-1 Rev.0 5.9 5.9-1 and 5.9-2 Rev.O Tables 5.5-1 N/A Rev.7 Figure 5.2-1 through 5.2-4 N/A Rev.5 Figure 5.2-5 N/A Rev.O Figure 5.2-6 N/A Rev.7 Figure 5.2-7 N/A Rev.5 Figure 5.3-1 N/A Rev.7 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 through 6.3-6 Rev.7 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.0 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.O O Revision 7 xxviii
.~..... - - LIST OF EFFECTIVE PAGES DEFUELED SAFETY ANALYSIS REPORT Section Effective Pages Revision 7.0 7.0-1 Rev.3 7.1 7.1-1 Rev.4 7.1 7.1-2 Rev.3 1 7.1 7.1-3 and 7.1-4 Rev.6 7.1 7.1-5 Rev.3 7.2 7.2-1 and 7.2-2 Rev.3 7.2 7.2-3 through 7.2-6 Rev.0 l 7.3 7.3-1 and 7.3-2 Rev.3 l 7.3 7.3-3 Rev.O 7.4 7.4-1 Rev.O 7.5 7.5-1 Rev.4 l 7.6 7.6-1 Rev.3 8.0 8.0-1 Rev.0 9.0 9.0-1 Rev.0 (') ~ i l l 1 1 xxix Revision 7 i l .i
1.2 GENERAL PLANT DESCRIPTION %J The Trojan Nuclear Plant site consists of approximately 623 acres located in Columbia County in NW Oregon on the Columbia River at River Mile 72.5 from the mouth. The distance from the reactor site to the nearest site boundary on land is 2172 ft. Major structures on the site include the Containment, Turbine Building, Auxiliary Building, Fuel Building, Control Building, and a single natural draft cooling tower. The town of St. Helens, Oregon, the county seat of Columbia County, is located approximately 12 miles SSE of the site. The town of Rainier, Oregon, is located approximately 4 miles NNW and the town of Kalama, Washington, is approximately 3 miles SE of the site. There are three small unincorporated communities within a 5-mile radius of the site: Prescott, Oregon, located approximately 1/2 mile N of the site; Goble, Oregon, located approximately 1-1/2 miles SSE of the site; and Carrolls, Washington, located approximately 2-1/2 miles NNE of the site. 1.2.1 DESIGN CRITERIA The principal design criteria for the Trojan Nuclear Plant are those fundamental architectural and engineering design objectives established for the Facility. The basis for development and selection of the design criteria used in this Facility were those which: (a) provide protection of public health and safety, (b) provide for reliable and economic Facility performance, and (c) provide an attractive appearance. The essential systems and components of the Facility are designed to enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena. The designs are based on the most severe of the natural phenomena recorded for the vicinity of the site, with margin to account for uncertainties in historical data. Q l.2-1
1.2.2 FUEL HANDLING SYSTEM i The Facility was designed to handle spent fuel under water from the time it leaves the reactor vessel until it is placed in casks for shipment from the site, although PGE is prohibited from moving fuel into the containment. Underwater transfer of spent fuel provides an optically transparent radiation shield, as well as a reliable source of coolant for removal of decay heat. ) 1.2.3 RADIOACTIVE WASTE TREATMENT SYSTEMS The radioactive waste treatment systems provide equipment necessary to collect, process, monitor, and discharge radioactive liquid, gaseous and solid wastes that are produced at the Facility. Liquid wastes potentially containing radioactive material are collected and monitored. Prior to l discharge, equipment is provided for filtering the liquid as required. Also, connections exist l for a demineralizer if filtration alone is not sufficient to meet discharge requirements. The treated water front the filters or demineralizers may be recycled for use in the Facility or may be discharged to the Columbia River. The miscellaneous dry waste, spent demineralizer resins and spent filters are shipped from the site for ultimate disposal in an authorized location. Gaseous wastes are collected and discharged to the environment after filtration to keep the offsite dose within prescribed limits. Revision 7 1.2 2
O :with the intake structure. Medium draft vessels would need a river stage o to clear the bottom while shallow draft vessels could clear the bottom during normal flow. In the unlikely event that all these conditions are met, the vessel, upon reaching the structure, would . need to either break through the walls and floors of the structure to damage the three service water pumps or cause sufficient damage to block the entire flow of water in the intake channel. Pleasure craft, because of the hull materials and weight could not damage the intake structure. Should the intake structure be sufficiently damaged by a ship collision as to be completely inoperable or the intake channel become blocked such that no water can be pumped, provisions l - have been made to maintain SFP makeup capability. [ 2.2.3 3 Oil or Corrosive Liould Spills in River Oli or corrosive liquid spills in the Columbia River could come from four potential sources: a ship or barge either in the ship channel, docked or grounded, a rail tank car leaving the tracks on s ) a bridge or the river bank, a truck leaving the road at a bridge or river bank, and industries on the river Provisions in the intake structure prevent floating petroleum products from entering the Plant systems. A concrete curtain wall extends to foot MSL which is below the lower river level, thus trapping oil at the surface. The chemicals associated with the Kalama plant are basically insoluble and would present no hazard. Other industries upriver (Portland and Vancouver) would have to release extremely large amounts of chemicals to have any measurable effect on the river at Trojan due to the large amount of river flow available for dilution. Using conservative values for diffusion coefficients, the effect on the river at Trojan as a result of dumping a ton of soluble chemicals in at Portland would be less than one part per million. A spill from trucks or rail tank cars would cause no hazard due to the smaller amount of chemicals involved and the dilution of the river. 2.2-21 Revision 7 i
A ship in midehannel would cause no significant hazard as the chemicals would be flushed past the site except for a small highly diluted amount. If a ship were to go aground in the intake channel and spill a large amount of chemical, some minor damage could occur to the externals of the intake structure. The Plant can be operated with no service water for a short time and spent fuel cooling can be maintained with no service water supply. 2.2.3.6 Cooling Tower Collapse The cooling tower is designed to withstand winds of up to 190 mph and earthquake loads of 0.15 g. In the unlikely event of collapse, the hyperbolic design of the structure, coupled with its thin-shell configuration, provide an inherently safe failure characteristic. The structure would tend to collapse inwardly. In addition, the structure is located sufficient distance (over 400 feet) from any equipment or structure important to safety, to prevent damage. O 2.2-22 1 i
4 2.3.2.1.2 Site Meteorological Data Meteorological data at the Trojan site during the period September 1,1971 through August 31, l 1974, are reported in this section(" "). These data compare favorably with national Weather Service data for Portland, Oregon. l 2.3.2.1L2.1 Average wind direction and sneed. The distribution of wind direction and speed is an important factor when considering transport conditions relevant to site diffusion L climatology. The topographic features of the site region are a major factor in influencing the wind direction distribution at Trojan. For the period September 1,1971 through August 31,1972, the prevailing wind for the 30-foot level is from the south, and south-southeast for the 200-foot level. For this period the average wind speed at the 1 30-foot level onsite was 8.2 mph and was 9.3 mph at Portland. s 2.3.2.1.2.2 Wind direction oersistence. Wind persistence is extremely important when considering potential effects from a contaminant release. Wind persistence is defined as a ' continuous flow from a given direction or range of directions. There is only a- ) 5-percent probability of continuous wind direction persistence periods greater than ) l 11.5 hours .ni, o2 l 2.3.2.1.2.3 Temocrature. humidity and precipitation. Temperatures in the Trojan region are generally mild considering its high latitude.The annual temperature onsite was 50 F. The l ' daily annual mean minimum onsite was 44*F and the annual mean maximum was 58*F Relative humidity is extremely important when considering the possible formation of fog in the site area in relation to the cooling tower. The mean relative humidity at Trojan is i 78 percent. Annual average precipitation in the Trojan region is 62.04 inches onsite. i 2.3-3 Revision 7 i'
2.3.2.1.2.4 Natt',al fog. Although the physical conditions that lead to fog are generally the same, the meteorological conditions may vary widely. In a previous study it was determined that fogs are more probable at night during stable or neutral conditions. In addition, the greatest frequency of occurrence of fog is during the fall season at night"*. 2.3.2.2 PotentialInfluence of the Plant and its Facilities on Local Meteorology Operation of the Trojan Nuclear Plant is not expected to affect the climate of the region. An evaluatien of the possible :nvironmental effects from the operation of a natural draft cooling towe,.a Trojan indicated the likelihood of any adverse effect from the cooling tower plume is extremely small"8 However, the physical structure of the cooling tower is expected to locally increase atmospheric turbulence. There is also a potential for somewhat decreased low-level wind speeds in the vicinity of the tower. This effect diminishes rapidly with increasing distance downwind from the cooling tower and is relatively insignificant offsite. O 2.3.2.2.1 Topographic Description General topography within a 50-mile radius is shown in Figure 2.3-19. The Trojan Nuc: ear Plant is located in the Columbia River Valley, which at this location is in a general north-south orientation. North of the Plant site the Columbia River bends to the northwest, and south of l the site the river bends to the southeast. Within the immediate vicinity of the Plant site there ( is a bluff one-quarter mile to the west rising sharply to 400-500 feet. North of the containment area there is a wooded hill which rises to 100 feet. The remaining area in the immediate l vicinity of the Plant site is flat and low. The Columbia River Valley is approximately 2 miles wide at the Plant site and widens to 3 miles north of the site at Longview-Kelse The valley walls at the h...: site rise to an elevation of 1000 feet MSL within approximately 1.8 miles to the west and not quite so high to the east. 2.3-4
1 The effect of the topographic features on airflow trajectory regimes and dilution is quite significant at Trojan. Analyses of annual wind roses reveal that the predominant wind flow is in a north-south direction. Winds within the Columbia River Valley will be effectively channeled and therefore will follow the changing orientations of this Valley. Computations of average x/Q values based on the straight line model for a ground-level release indicate that the l greatest potential concentrations would be north and somh of the site, corresponding to the predominant wind directions. It should be emphasized that the assumption of a ground-level release is conservative and results in maximum ground-level concentrations which are generally not sens!!ive to a change in elevation of the terrain. In addition, a nonbuoyant plume will generally not rise out of the valley for a ground-level release during stable temperature lapse rate conditions.' Estimates of dispersion during stable conditions, based on the Gausian i l diffusion model, indicate that a plume oriented in a general north-south direction would most l likely not intersect with the valley walls. Therefore, the valley walls have only a limited eifect i as a potential barrier to prevent dispersion of the plume since the width of the valley increases - both to the north and south of the containment site and the plume width is relatively narrow during stable conditions. Turbulence created by the mountainous terrain would increase the dilution of airborne effluents. i l 2.3.2.3 Local meteorological Conditions for j Design and Ooeration Bases l No local meteorological conditions have been identified that are used for design and operating bases, other than those identified in Section 2.3.4. Regional meteorological conditions used l for design and operation bases are discussed in Section 2.3.1.2. 2.3.3 ONSITE MFJEOROLOGICAL MEASUREMENTS PROGRAM l Permanent plant defueling has resulted in significant reduction in source terms. DBAs do not l h; result in releases which would exceed 10 CFR 100 limits. The reliance on meteorological l 2.3-5 Revision 7 L.
l monitoring for the calculation of off-site doses from normal operation and accident conditions j has essentially been eliminated. Therefore, the on-site meteorological program has been l terminated. l 2.3.4 DIFFUSION ESTIMATES Diffusion estimates were made for both short-term and bng-term conditions. The development for each of these estimates is provide 4 below. For the short-term diffusion estimate a hypothetical accident was postulated to determine the concentrations and dosages that might occur in the event of a release for the following time periods: I hour,2 hours,8 hours,16 hours,3 days, and 26 days. A basic input for the j calculation is the onsite meteorological data (September 1,1931 through August 31,1972) I o2x which determines the dilution capacity of the atmosphere 0 The short-term diffusior. estimates were obtained by the following equation: i x 1 p (no, o,1 0, m u.1 +cA)p, (2.3-5) y where x = concentration, units /m' Q = source strength p = mean wind speed, m/sec Revision 7 2.3-6
l o = horizontal dispersion parameter, m y U o, = vertical dispersion parameter, m c = building shape factor, dimensionless (0.5) A = smallest cross-sectional area of the containment 2 Structure,2340 m m = the time allotted for the running average, hour k = hour index j = sector ofinterest. y3 ) The x/Q values are presented in Tables 2.3-1,2.3-2,2.3-4, and Figure 2.3-1. Annual average atmospheric dilution factors (x/Q), were conservatively calculated for the Trojan site based on onsite data for the period September 1,1972 through August 31,1974 by assuming ground level releases using Equation 2.3-6. This equation assumes a horizontal distribution within a 22.5 degree sector. Stability is based on AT data. Calms were distributed based on the directional frequency of winds in the 0.6 to 1.5 mph range and were assigned a wind speed of 0.3 mph. Limited vertical mixing also was accounted for due to the mixing depth which was taken as an average of 1000 meters for the Trojan site. Calculations of annual average X Q values then were adjusted for temporal / variation in the air flow of the site. (5)2 { ### ( 5 ),yu] = O 2">" N O (2.3.6) (m, C) 2.3-7 Revision 7
where (5)'#* if a*>>0.465L = 0 ( E,, u,x) (2.3-7) 2hL* ( 5 ) jj,= (E,j,x) 3.of(ah)exp (2.3-8) 2.032 0 u Ah = 1, h = 0 2, h > 0 if 0.465L s o,3 < 1.6L .546 (5)21* if 1.6L < a'1 = 0 ( Lu,x) (2.3-9) O and (x/Q), = relative ground-level concentration normalized by Qu source strength Q for sector k, sec/ meter' (X Q)jt = relative ground-level concentration x normalized by Q,;, source strength Q / for stability class j, wind speed class i, and wind direction sector k, sec/ meter' 2 [,, = ( o,, O. 5 H ) m 2 n Revision 7 2.3-8
b i with the constraint that A 2) zj l N.= total observations for ' data period l n, = total observations for stability class j, wind speed class i and wind direction sector k o = vertical' stability parameter for stability class j, meters g l H = the height of the containment, meters L = the mixing height,1000 meters x = the downwind disuun:.. c;ters t u, = midpoint wind speed of wind speed class i, meters /sec. The values of o are based on curves presented in Regulatory Guide 1.111. g Calculations of annual average D/Q values were made as follows: 8 df* i ( )"= nox (2.3-10) l 2.3-9 Revision 7
where h (D/Q), = relative deposition per unit area for sector k, m.2 x = downwind distance, miles d/Q = relative deposition rate per unit downwind distance, m f, = relative frequency of wind direction into sector k, dimensionless. The values of d/Q are based on the curves for ground-level releases of Regulatory Guide 1.111. Since the straight-line flow model does not consider the temporal variation in the airflow of the site region, terrain adjustment factors for the Trojan site were developed from a segmented plume model (NUSPUF) calculation based on 10-min. averages of wind and AT data, a straight-line model (Equations 2.3-6) and the methodology of Reference 28. These values are based on AT m feet-30 feet stability data and 30-foot wind data for the period August 1,1976 2 through July 31,1977. Terrain adjustment factors were determined for downwind distances to 5 miles north and south of the site and to distances of 3 miles west and 3.5 miles east. For distances beyond the area of analysis or for those distances where the model indicated that the terrain adjustment factcr would be less than 1.0, the terrain adjustment factor was conservatively set to 1.0 to adjust X Q and D/Q values presented in Tables 2.3-2 through / 2.3-4. Tables 2.3-5 and 2.3-6 present the maximum terrain adjustment factors for each annular sector for the standard population distances and for the special distances. Revision 7 2.3-10
m_. i 2.4.6 PROBABI E MAXIMUM TSUNAMI FLOODING Historically the evidence demonstrates that the mouth of the Columbia River is relatively 1 - insensitive to tsunamis when compared to Crescent City, California, 310 miles south of the 1 Columbia River entrance. The tsunami effects at the mouth of te Columbia River are further dissipated inside the river due to the characteristics of the estuary, as was demonstrated during the tsunami generated by the Alaskan earthquake of March 28,1964. The elevation of the Plant site is 45 feet MSL. Because of the large margin between the Plant elevation and the river surface and because of the insensitivity of the river to tsunami effects, tsunamis are not considered in the design criteria for the Trojan Plant, s 2.4.7 ICE EFFECTS The general climate in the lower Columbia River Basin is not conducive to ice formation. In j addition, the flow of the river during periods of freezing temperatures is sufficiently large I (200,000 to 400,000 cfs) that ice formation is impossible in the main streamflow. During extended periods of freezing temperatures, some icing is experienced along the banks of sloughs and inlets where the water is slow moving or stagnant. l Any surface ice formation in the vicinity of the intake structure would not affect the operation of the intake structure, as the water entrance is below the water surface. The lowest recorded river temperature at the site was 34.1*F on February 6,1971 (period of record 1967 to 1972)"".
- O 2.4-31
\\. i
2.4.8 COOLING WATER CANALS AND RESERVOIRS The cooling water for the Trojan Plant is drawn in through an intake structure located on the bank of the Columbia River and oriented perpendicular to the flow of the river. The intake channel provides a water channel to the intake structure during low flows. During j a flood, the intake channel would be kept clear by the scouring action of the water flow. l [ 2.4.9 CHANNEL DIVERSIONS \\ t The Columbia River, for most of its lower reach, flows through a relatively narrow gorge. Diversion or rerouting of the river is impossible because of this gorge and the mountain ranges through which the river flows. Blockage of flow due to a catastrophic landslide or some other incredible event upstream of the Plant would not present a hazard. Should flow in the river reach zero (which would require blockage of all tributaries also), the river channel would act as a extension of the Columbia River estuary. The service water pump inlets are located at -8 feet MSL. i 2.4.10 FLOODING PROTECTION REOUIREMENTS l All facilities / equipment required for the storage of irradiated fuel are located above or are protected for water levels to Elevation 45 feet MSL. The intake stnicture is the only structure l exposed to the river flow below Elevation 45 feet MSL. The unlikely event of an intake l structure failure could result in the loss of service water makeup supply to the SFP. At least l 10 days are available after the loss of both forced cooling and makeup to the SFP to establish l makeup flow to the SFP and maintain required cooling. Revision 7 2.4-32
2.4.11 LOW WATER CONSIDERATIONS Low water in the Columbia River does not affect the safety of the storage of irradiated fuel. The water intake for the Plant (service water pumps) is located below mean sea level, thus providing water to the Plant under any conditions of low flow in the Columbia River. The SWS can provide adequate water for SFP makeup with the river surface as low as -2.0 feet l MSL. Due to the large amount of storage on the Columbia River system, and to the extent of the Columbia River Basin, the minimum daily average flow of the river expected is 110,000 cfs"8). The minimum mean daily flow recorded at the site was 94,000 cfs on April 18, 1968, when the flow was deliberately regulated at Bonneville Dam to determine the downstream t effects of low flow. Surge-or tsunami-caused low water is not applicable to the Trojan site. O V 2.4.12 DISPERSION, DILUTION, AND TRAVEL TIMES OF ACCIDENTAL RELEASES OF LIOUID EFFLUENTS IN SURFACE WATERS . Effluent from the Trojan Plant enters the Columbia River through the outfall line. This line is a 3-foot-diameter pipe set perpendicular to the river flow near the river bottom. Discharge from the pipe is through 2401-1/2-inch-diameter ports. The ports are horizontally aligned at 1-foot intervals on both sides of the pipe for the last 120 feet of the pipe. The highest port is approximately -31 feet MSL and the lowest at approximately -49 feet MSL. The outfall diffuser has been designed for maximum dispersion and dilution of the effluent before it reaches the river surface. en 2.4-33 Revision 7
1 The dispersion and dilution of the Plant outflow has been studied for both short-and long-range effects. Studies conclude that the minimum dilution ratio to be expected at the water surface above the pipe is greater than 100, based on worst-case assumptions"*. To evaluate the long-range effects on the effluent, PGE engaged Battelle Memorial Institute Pacific Northwest LaboratoryG 2". The results of their dye tracer studies indicate that the continuous discharge flow from Trojan would be completely mixed with the river flow within 12,000 feet downstream of the discharge. The nearest surface water use downstream of the outfall line is at the Rainier municipal water intake, 4.5 miles downstream. s Additional studies have been made to determine the impact of a block release on the intake for the Rainier water supply. Diffusion coefficients for the river were both measured and calculated from theoretical equations. The theoretical coefficients were used in the evaluation for conservatism". 2.4.13 GROUNDWATER 2.4.13.1 Descriotion and Onsite Use The Trojan site is located on an extremely impervious rocky ridge that is bounded on one side and end by the Columbia River and on the other side and end by an old river channel that has been completely filled with alluvial sediments. i The slough is arcuate, about 2,000 feet wide and 2 miles long. It is carved in the Eocene Goble Volcanics, which borders the slough on the west side, and forms a rock knob on the east side that separates the slough from the Columbia River. The ground elevation O 2.4-34
4 2.5 GEOLOGY. SEISMOLOGY AND GEOTECHNICAL ENGINEERING 'In compliance with the criteria provided in Appendix A, " Seismic and Geologic Siting Criteria for [ . Nuclear Power Plants", of 10 CFR 100 and the new NRC standard format, this section describes. ~ and evaluates the geolhgic and seismic conditions for the region around the Trojan Plant site. l Foundation conditiom are evaluated, and the foundation design is described. The seismic history of the region is examined, and the earthquake design criteria are developed and described. 2.5.1 BASIC GEOLOGIC AND SEISMIC INFORMATION i i j' The Trojan Plant site is 31 miles north of the city limits of Portland, Oregon on the Oregon bank of the Columbia River. A portion of the site is underlain by a north-south trending steep-sided i ridge of volcanic rock that borders the river and rises to a maximum elevation of 134 feet above MSL. All Seismic Category I structures and Seismic Category II structures housing Seismic Category I equipment are founded on this rock ridge, except two footings for the diesel generator l. rooms which are placed on compacted backfill.. The remainder of the site is underlain by a flat alluvian plain with elevation ranging between 5 and 18 feet. Approximately 1/2 mile west of the site, a north-south trending range of hills rises steeply above the alluvial plain to elevation in i excess of 1000 feet MSL. The Columbia River flows in a northerly direction at the site but turns to the west several miles downstream. 1 i; I . Investigations were conducted to determine the characteristics of the foundation material, especially l in regard to their suitability for supporting the structures, to determine the depth and configuration of the groundwater table, to determine the characteristics of the soil and rock materials with , respect to their effect on the migration of radioactive solutions if such solutions come in contact with them, and to evaluate the seismicity of the area so that appropriate parameters for seismic design could be selected. O 2.5-1 Revision 7
Consultants in geology and seismology were retained to evaluate independently the results of the field investigations. A river bottom survey (" Boomer" survey) was made by EG&G of Goleta, California, using continuous seismic profiling, to define the shape of the river bottom adjacent to the site. A geophysical survey was performed by Geo-Recon Inc., of Seattle, Washington, across the alluvial valley to the west of the site. A geophysical survey was made at the reactor site, by P.C. Exploration of Carmichael, California, to meace P-wave and S-wave velocities and to calculate the dynamic modulus of elasticity of the foundatiori rock. Drilling and sampling was done by Lynch Bros. of Seattle, Washington, under the direction of Bechtel Corporation. Soil tests were designated by Bechtel and done by Shannon and Wilson, Inc., Seattle, Washington, in their Portland laboratory. Selected rock core samples from the drill holes were tested by Bechtel in l their laboratory in San Francisco. A comprehensive geophysical survey was made to investigate geologic conditions in the Columbia River channel adjacent to the site. Studies included seismic refraction, resistivity, aeromagnetic, and gravity surveys. The results of these investigations were evaluated by a special advisory board and the results are presented in a geophysical survey report dated August 1,1972m, 2.5.1.1 Regional Geology l l l The Trojan Plant site is located in the Oregon Coast Range section of the Pacific border physiographic province. The Coast Range is farther divided, and the site is on the southeastern margin of the Willapa Hills subsection, The Coast Range section is bordered on the north by the l Olympic Range and on the south by the Klamath Mountains. In the area near the site and along the i northern two-thirds of the Coast Range, the Puget Trough forms the eastern boundary. The southern third is bounded on the east by the Sierra-Cascade Province. O 2.5-2
~ - -. -.. . - _.. - ~.. - - (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 area,m (8) A fuel 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 4 The SFP is a reinforced concrete structure with seam-welded stainless steel plate liners. The pool . volume is approximately 51,900 ft' with a surface area of 1300 ft. The pool is filled with borated 2 water which is maintained at a concentration of 22000 ppm boron. The spent fuel assemblies are stored in storage racks in parallel rows having a center-to-center distance of 10.5 inches in both horizontal directions. The 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 pro'vided 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 adjacent pools drained, water pressure tends to seal the doors. Additionally, 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 ' ater level in the SFP is maintained to provide at least 23 feet of water above the top of a spent w fuel assembly in the storage racks, and at least 9.5 feet above the active portion of the fuel assembly O O 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. g 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 l liner. The trenches drain through normally open valves to the Liquid Radioactive Waste Treatment l System. The leak detection valves are arranged into manifolds that are inspected periodically to 1 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 (PEPMS) 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 Evaluatio.n The center-to-center distance between the adjacent spent fuel assemblies and the fixed neutron absorber are sufficient to ensure a k, s0.95 even if unborated water and fresh nondepleted fuel, enriched in U-235 to 4.5 wt%, are in the SFP. Revision 7 3.2-38
l l l l 3.3 AUXILIARY SYSTEMS l l This section discusses the auxiliary systems that are used to support the storage of spent fuel at Trojan. This section includes discussions on the Fuel Handling System, Modular SFP Cooling l and Cleanup System, Service Water System, Compressed Air System, Makeup Water Treatment l 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. l 3.3.1 FUEL HANDLING SYSTEM l - 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 L prohibited under Trojan's current license. O 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 l design bases apply to the Fuel Handling System: l (1) Fuel handling devices have provisions to avoid dropping orjamming of fuel I assemblies during transfer operations. l i O ) 3.3-1 Revision 7
l l l (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 l cause damage to any fuel elements should a SSE occur duting fuel transfer operations. (3) The hoist used to lift the spent fuel assemblies has a limited maximum lift height which is determined by the length of the long-handled tool, so that the minimum required depth of water shielding is maintained. Environmental conditions of the fuel handling equipment, such as exposure to borated water and high humidity, are considered in the design and selection of the material. 3.3.1.2 System Descrintion 3.3.1.2.1 General Description Fuel assemblies are moved in the SFP using the SFP bridge hoist. When lifting spent fuel assemblies, the hoist uses a long-handled tool to assure that sufficient radiation shielding is maintained. Fuel assembly inserts, such as thimble plugs, burnable poisons rods, rod control clusters, and source rods, may also be transferred between positions within the SFP. i l 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 f 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 3.3-2
t 3.3.1.3.2 Radiation Shielding During spent fuel transfer, the gamma dose rate at the surface of the water is 2.5 mrem /hr or less. This is accomplished by maintaining a minimum of 9.5 feet of water above the top of the active l portion of the fuel assembly during handling operations. The hoist on the SFP bridge crane moves spent fuel assemblies with a long-handled tool. Hoist travel and tool length limit the maximum lift i l of a fuel assembly to within 9.5 feet of the normal water level in the SFP. l l 3.3.2 MODULAR SFP COOLING AND CLEANUP SYSTEM l i l The Modular SFP Cooling and Cleanup System, shown in Figure 3.3-4, removes the decay heat l from the spent fuel elements stored in the SFP and purifies the system water inventory. The System does not perform any safety functions. i l 3.3.2.1 Design Bases ,lO The Modular SFP Cooling and Cleanup System is designed to perform the following functions: l l (1) Maintain the SFP water s 140 F with the heat load associated with the 781 spent fuel l assemblies stored in the SFP after five years following the 1993 permanent shutdown l and defueling of the Plant l (2) Maintain sufficient cooling of fuel assemblies in the event a fuel assembly or other object is dropped and remains lying across the top of one or more assembly locations (3) Maintain clarity and purity of the borated water in the SFP l l [G lG 3.3-5 Revision 7
l Decay heat values for evaluating the SFP and the original cooling system capabilities were derived utilizing ANSI /ANS 5.1-1979m. The thermal analyses included the effects of manufacturing tolerances of the racks and uncertainties of fuel assembly positions in the rack. The effects of gamma heating on SFP wall and floor temperatures were evaluated to verify that acceptable temperatures are maintained. l The Modular SFP Cooling and Cleanup System is designed to Seismic Category II requirements. 3.3.2.2 System Description l The Modular SFP Cooling and Cleanup System is shown in Figure 3.3-4. The system consists of the following components: l (1) Two full capacity modular cooling pumps ] (2) Two water-to-air coolers, each consisting of a fan and cooling coil, capable of h l removing more than 870,000 BTU /hr at 55"F outside air temperature and 110 F pool l temperature l (3) One modular SFP purification filter l l (4) One modular SFP demineralizer l (5) One SFP skimmer l (6) Valves, hoses, and piping l (7) Instrumentation i O l Revision 7 3.3-6 l i i
The Modular SFP Cooling and Cleanup System components are located in the Fuel Building l r~ y} adjacent to the SFP. l The Modular SFP Cooling and Cleanup System is a water-to-air closed-loop system. It takes j suction from the SFP to one of the two cooling pumps installed in parallel. The cooling pump l discharge may be aligned through the purificmion filter and/or the demineralizer and back to the l pool. The system is capable of a flowrate through the demineralizer sufficient to recirculate the SFP l inventory approximately once every two days. This purification loop is ndequate for removing l fission products and other contaminants which may be introduced by a leaking fuel assembly. Portable vacuum filtration units may also be used to support special cleanup evolutions. l l From the discharge of the demineralizer the water is routed to one or both of the air coolers installed l in parallel. The air coolers are provided with outside air through an inlet louver located in the east l wall of the Fuel Building. The cooling air passes through the air cooler (s) and discharges back to l the outside through a discharge louver in the south wall of the Fuel Building. An enclosure l g T separates the outside cooling air from the Fuel Building inside air. A radiation detector is mounted l V on the drain piping from the air cooler enclosure to detect air cooler coil leakage. l l Sample points are provided at the inlet and outlet of the demineralizer to evaluate its performance l and to monitor SFP chemistry. Table 4.3-1 lists the SFP chemistry specifications and sampling l requirements. l l A SFP floating skimmer is connected to the suction piping via a short hose. A small fraction of the l water entering the suction piping is drawn from the top of the pool surface through the skimmer. l O 3.3-7 Revision 7
} 3.3.2.3 Design Evaluation l A single modular SFP cooling pump and air cooler can maintain SFP temperatures s 140 F under l all anticipated climatic conditions. l The capacity of the Modular SFP Cooling and Cleanup System is based on the heat load with a l maximum of 781 fuel assemblies stored in the SFP after five years following the 1993 permanent l shutdown and defueling of the Plant. In the current permanently defueled Plant condition, no l additional fuel assemblies will be stored in the SFP. l l The ends of the suction and discharge pipes only extend approximately two feet below the normal l pool water level elevation of 91 feet. This elevation (88 feet,10 inches) is well above the 83 feet, l 11 inch elevation of the siphon breakers for other piping entering the SFP. Thus the accident l analysis for a loss of SFP inventory due to a siphoning event is bounded by the analysis in l Section 6.3. O l The Modular SFP Cooling and Cleanup System does not perform any safety functions. Loss of forced SFP cooling will cause the SFP water temperature to slowly rise. The longest time interval between SFP cooling system failure and its detection is 24 hours since the systern is inspected once per shift. The failure would be detected sooner if the SFP water temperature reaches the alarm setpoint. If forced cooling cannot be restored, then the SFP water temperature will continue to rise, increasing the evaporation rate and possibly resulting in boiling within the SFP. The only requirement to assure adequate cooling of the spent fuel is to maintain the water level in the SFP above the spent fuel elements. Section 6.3 analyzed the loss of forced cooling to the SFP and determined that sufficient time exists to effect repairs to the cooling system or to establish makeup flow prior to uncovering the spent fuel. Makeup water is available from a variety of sources as described in Section 4.3.1. Revision 7 3.3-8
s In the event of a design basis seismic event, the systems that were designed to Seismic Category II requirements may not be operable. In this case, sufficient time would be available to align a source of makeup water to maiutain water level. n 3.3.3 Deleted l 3.3.4 SERVICE WATER SYSTEM \\ The SWS is designed to provide water from the Columbia River to supply water to various systems l-and equipment. With the reactor permanently defueled, the primary function of the SWS is reduced l 4 to providing makeup to the SFP. l 3.3.4.1 Design Bases The SWS is designed to deliver the minimum required flows of water to equipment assuming a .O's.J minimum water level of 1.5 feet below MSL in the Columbia River. With the reactor pe defueled, system design requirements are reduced substantially from the original design bases of the system. The system design includes provisions for inhibiting long-term corrosion and organic fouling of the system water passages. 3.3-9 Revision 7
3.3.4.2 System Description O l The components included in the Service Water System are: l (1) Three SWS pumps (P-108A, B, and C). Note that P-108A has been rebuilt. It may l remain removed from the system indefinitely l (2) One SWS strainer (F-101B) l (3) Interconnecting piping, valves, and instrumentation l (4) Two cooling water pumps (P-167A and B) 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. O One traveling water screen is installed in the flow path to each of the two independent flow paths to the service water pumps at the intake structure. The screens are automatically cleaned by the screen wash system which consists of two vertical pumps taking suction on the downstream side of the screens and discharging into high velocity spray nozzles which clean the debris from the screens as they travel past the nozzles. The screen wash system is automatically actuated by an adjustable timer or when there is a high differential level across the screen. Three identical service water pumps take suction from the river through the traveling water screens at the intake structure. Each pump was designed to provide 100 percent of the operating Plant flow l requirements. Power to P-108A 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. Lubricating water is supplied to the three service water pumps. River Revision 7 3.3-10
- -.-. = - - water can also be supplied to the SWS using Cooling Water Pumps P-167A and P-167B. P-167A j r\\ (Q can supply makeup flow to the SFP. l Two identical service water strainers are provided for straining the discharge from the service water pumps and the cooling water pumps. One strainer is located in the supply header to each loop. Two identical service water booster pumps, P-148B and D, are provided to boost the pressure in the i water supply lines to components served by the system except the supplies to the CCWS heat exchangers, bearing lube water and SWS strainer backwash. These are supplied directly from the l service water pumps. Either service water booster pump is capable of providing in excess of 100 percent of the loop flow requirements. l For the defueled condition, train independence and automatic isolation of the Seismic Category II loads are not required. A single SW loop provides excess capacity for the current requirements. l 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. l 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. O 3.3-11 Revision 7
l-The cystem 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 h unit connects to the air-dryer unit inlet and the service air header. The instrument air header is connected to the air-dryer unit discharge. Each air header supplies branch lines which supply instrument air and service air to the required loads throughout the Plant. The instrument and service l air system provides air to the inflatable seals for the SFP gates. 3.3.6 BORIC ACID BATCH TANK The boric acid batch tank will normally be used to supply borated water to the SFP. Procedural controls will be used for this process. 3.3.7 MAKEUP WATER TREATMENT SYSTEM i l The Makeup Water Treatment System provides demineralized water of the required quality to meet Plant needs. Makeup water is processed and then stored in the demineralized water storage tank l where it is available to meet Plant needs. 3.3.8 EOUIPMENT AND FLOOR DRAIN SYSTEMS i l The following equipment and floor drainage systems are provided for the Plant: l l (1) Liquid Radioactive Waste Treatment System (LRWS) drains l l (2) Oily Waste System i l (3) Acid Waste System l (4) Sanitary Waste System O1 Revision 7 3.3-12 l
4 l The equipment and floor drain systems do not perform any safety functions. I The LRWS is designed to collect liquid waste from areas containing equipment that handles l . radioactive fluids. This system is designed to control the spread and release of radioactive l particulates by directing potentially radioactive fluids to the Radioactive Waste Treatment System. l l This system is described in Section 5.3. l t . The Oily Waste System collects the waste from floor and equipment drains in places where the i 4-waste is not potentially radioactive. The waste is conveyed to a settling tank where the oil is separated prior to releasing the water to the Plant discharge and di! : tion struchtre. t The Acid Waste System is designed to drain fixtures and equipment in which chemicals are expected to be present in nonradioactive effluent. Selection of piping materials was based on providing a surface resistant to corrosion. The waste is drained into the neutralizing tank, T-126, where it is neutralized. The neutralized waste is then transferred to '.he 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 connections, and the plumbing fixtures. 3.3.9 PLANT DISCHARGE AND DILUTION STRUCTURE The Plant discharge and dilution structure receives the Plant liquid radioactive and chemical wastes, provides dilution water, and discharges the diluted waste to the Columbia River. The residual chlorine concentration of the discharge is controlled by the addition of sodium bisulfite at the Plant effluent. Diluted chemical waste discherge concentrations are diluted to meet the requirements of . the State of Oregon Department of Environmental Quality. The structure does~not perform any safety functions. O 3.3-13 Revision 7
l 3.3.10 Deleted 3.3.11 FIRE PROTECTION SYSTEM AND PROGRAM Fire Protection is provided for the Plant as described in the Topical Report, " Trojan Nuclear Plant Fire Protection Program", PGE-1012. 3.3.12 CONTROL ROOM HABITABILITY To support habitability, the original Control Room design included radiation shielding, air filtering, air conditioning and ventilation systems, fire protection, personnel protective equipment and first aid, and utility and sanitary facilities. As discussed in chapter 6, the only accident requiring operator action is a prolonged loss of SFP cooling. Since this postulated event does not require operator action for several days, short term actions initiated from the Control Room to restore SFP cooling or to establish SFP makeup water flow are not required to protect the health and safety of the public. Those systems originally provided to assure habitability during accidents are no longer required. 3.3.13 SEISMIC INSTRUMENTATION Seismic instrumentation for the facility consists of a multielement seismoscope and peak acceleration recorders. The multielement seismoscope is mounted on an essentially rigid foundation, which will provide no significant amplification of ground motion. Peak acceleration recorders are also installed at appropriate locations in the facility. The seismic instrumentation satisfies 10 CFR 100, Appendix A, which requires instrumentation so that the seismic response of features important to safety can be determined promptly to permit comparison of such response with i that used as the design basis. Revision 7 3.3-14 l
c T v) 'J 5 TABLE 3.2-4 CAI.CUlJLTED RESUI.TS - CONTROL Bull.DINC - - - ~ ~ - - - - - y t4actlption Load Strees calculated A11ovable of Healer tacation of Hesber Combination Combination Strees Strees Reserke W 24 m 76 sean floor of control rneo on Col. Line O D + 1. 18.8 kol 24.0 kel be t we e n 4 6 a nd 51 - El. 93 '-O" I '-0" Conc r e t e Floor of control reos on Col. I.ine 0 1.5 D + 1.8 1. 7.5 k-ft 8.6 k-ft Alloweble la Ultimate Stab between 46 and 51 - El. 9)*-O" Homent Cepecity W IS x 45 Seen Floor of cable spreading room on Col. Lane O D+L 16.0 kel 24.0 kel between 46 and 51 - El. 77'-0" 6" Concrete Floor of cable spreading room on Col. Line 0 1.5 D + 1.8 L 1.8 k-ft 5.6 k-ft Allowable la Ultimate Sieb between 46 and 51 - El. 77'-0" Homent capeetty W 14 s 84 Column Co lumn I.i ne 0 - 4 6 - El. 6 t '-0" D+L 0.942 1.0 Alloweble le by Column Interaction Formule W 12 m 142 Column Coluen Line 0 21. 61'-0* O.t L 0.979 1.0 A13ewable is by Cclumn Internetten Formule foote s From the indicated results it is apparent that straine are not criticalg therefore, strain values are not given. o$. g. n
TABLE 3.2-5 i SIGNIFICANT ELEVATIONS IN THE SPENT FUEL POOL Description Elevation 93'0" Top of the spent fuel poo! Normal water level approximately 91' Ends of the suction and discharge pipes for 88'10" l the Modular SFP Cooling and Cleanup l System l Water level after loss of gates to the fuel 84' transfer canal and cask loading pit with the cask loading pit filled Siphon breakers 83'l1" Bottom of gate to the fuel transfer canal 68'4" 67'4" Top of fuel racks Top of fuel assemblies 66'7" Wetted side of the bottom of the spent fuel 52'6" pool O l Revision 7
TABLE 3.5-1 Sheet 3 of 38 ' Regulatory Guide L 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized l . Water Reactors (3/72), Rev. 0 l: l Compliance Status Coraply with exception. The assumptions used in this Regulatory Guide for fuel rod gap fractions and iodine pool decontamination factors are based on certain fuel performance limits. One of the assumed limits of assembly average burnup <25,000 mwd /MTU is lower than expected for Trojan l fuel. Specifically, Trojan fuel average burnups can be as high as l 44,000 mwd /MTU. However, an analysis showed that the gap fractions in the Regulatory Gitide still conservatively bound realistic case calculations up to 44,000 mwd /MTU. Therefore, the assumptions of this Regulatory Guide are still valid for ] Trojan fuel. Regulatory Guide v 1.26 - Quality Group Classifications and Standards for Water, Steam, and Radioactive-l - Waste 4ontaining Components of Nuclear Power Plants (2/76), Rev. 3 l Compliance Comply with exception. l Safety-related struct s, systems, and components were originally classified into quality groups prior to the suance of Revision 0 to Regulatory Guide 1.26. Currently, the facility I classification process is described in PGE-1052, " Quality-Related List Classification Criteria for the Trojan Nuclear Plant", for the permanently defueled condition. Regulatory Guide 1.28 - Quality Assurance Program Requirements (Design and Construction) (8/85), Rev.3 i } Revision 7 t
TABLE 3.5-1 Sheet 4 of 38 Comoliance Comply with exception. The Trojan Nuclear Plant was initially designed and constructed in accordance with the Quality Assurance Program included in Chapter 17 of the FSAR. This pregram contained those measures necessary to assure adequate quality in the completed facility in accordance with Appendix B to 10 CFR 50. Subsequent and future modifications to the Trojan Nuclear Plant are carried out under the provisions of Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operation), (Revision 2)" and the guidelines of PGE's Nuclear Quality Assurance Program (PGE-8010). Regulatory Guide 1.29 - Seismic Design Classification (9/78), Rev. 3 Comoliance Comply with exception. (a) Regulatory Positions C.1.d and C.1.g require that the cooling water and systems or portions of systems that are required for cooling the spent fuel storage pool be designed l as Seismic Category I. However, the Modular Spent Fuel Pool Cooling and Cleanup System has been designed to Seismic Category II requirements. In the current defueled condition, the reduced decay heat available to heat-up the spent fuel pool allows adequate time to supply water for makeup due to boiling from numerous non-seismic sources. This diversity ensures adequate protection of the spent fuel. (b) Footnote 2 to Regulatory Position C.1.p states that specific guidance on seismic requirements for radioactive waste management systems is under development. Seismic requirements for radioactive waste management systems are addressed in Regulatory Guide 1.143 (Rev.1 October 1979). Regulatory Guide 1.143 requires the foundations and walls of structures housing the liquid and solid radwaste systems to be Seismic Category I, and permits the equipment used to collect, process, and store liquid and solid radioactive wastes to be Seismic Category II. Regulatory Guide 1.143 also requires portions of gaseous radwaste treatment systems that are intended to store or delay the release of gaseous radioactive waste, including portions of structures housing the system, to be Seismic Category I. Whereas Trojan's seismic classification guidelines for radioactive waste management are not as specific as those given in Revision 7
l l 1 l TABLE 3.5-1 Sheet 5 of 38 b l Regulatory Guide 1.143, the seismic classification of Trojan's solid, liquid, and I gaseous radwaste systems, and the structures housing the systems, are in full compliance with the above seismic requirements of Regulatory Guide 1.143 for such systems. l (c) Regulatory Position C.1.p of Revision 3 to Regulatory Guide 1.29 also requires that l any other systems (other than radwaste systems) not covered specifically in the i Regulatory Guide that contain or may contain radioactive material, and whose postulated failure would result in conservatively calculated potential offsite doses greater than 0.5 rem (whole body or its equivalent to any part of the body), be designated Seismic Category I. The Trojan classification method does not contain this l additional requirement for the classification of structures, systems, and components outside of the Control, Auxiliaty, and Fuel Building Complex. These items were classified in accordance with Revision 0 of Regulatory Guide 1.29. (d) Regulatory Position C.2 of this Regulatory Guide requires that portions of non-safety-related systems whose failure could reduce the functioning of a Seismic Category I l plant feature to an unacceptable level be designed and constructed so that the SSE would not cause such failure. The original Plant design included system interaction L considerations as well as failure modes and effects analyses primarily in determining system and equipment locations. Portions of non-safety-related Seismic Category II systems (e.g., pipe supports) were not originally designed to Seismic Category I requirements. The system interaction requirements in this Regulatory Guide are now implemented in Seismi; II/I provisions. A system Literaction review identifying potential II/I items has been completed for safety-related systems and equipment in the Control, Auxiliary, and Fuel Buildings. Design and analysis of Seismic Category II/I items are for SSE seismic loadings (operating basis earthquake loads are not required te be analyzed). The design and construction of Seismic Category II/I items will be considered quality-related but not safety-related. The original Plant design meets Regulatory Position C.2 that is giver' in Revision 0 to this Regulatory Guide. The position to design and construct Seismic Category II/I items for SSE loads complies with Revision 3 to this Regulatory Guide, j l ,[O Revision 7 i l
l TABLE 3.5-1 Sheet 6 of 38 Regulatory Guide 1.30 - Qur.lity Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electrical Equipment (8/72), Rev. O Compliance Comply. Reculatorv Gride 1.33 - Quality Assurance Program Requirements (2/78), Rev. 2 Cnmpliaura Comply with exception. l (a) Section 4.3.1 of ANSI N18.7-1976 requires personnel assigned responsibility for l-independent reviews to collectively have the experience and competence required to l review problems in the following areas: nuclear power plant operations; nuclear l engineering; chemistry and radiochemistry; metallurgy; nondestructive testing; l instrumentation and control; radiological safety; mechanical and electrical engineering; j administrative controls and quality assurance practices; and other appropriate fields l associated with the unique characteristics of the nuclear power plant involved. l In lieu of the above, as specified in Technical Specification 5.5.2, personnel assigned l to the Independent Review and Audit Committee shall 'ollectively have experience and l knowledge in the following functional areas: fuel handling and storage; chemistry and l l radiochemistry; engineering; radiation protection; and quality assurance. l l (b) Section 4.3.4(2) of ANSI N18.7-1976 requires the independent review body to review l l " Proposed changes in procedures, proposed changes in the facility, or proposed tests or l experiments, any of which involves a change in the technical specifications..." PGE l l meets the intent of this requirement via the Independent Review and Audit Committee's l l review of the tssociated technical specification change. l l l (c) Section 5.1 of ANSI N18.7-1976 requires a summary document of those sources providing administGtive controls and quality assurance during the operational phase to index such source documents to the criteria of ANSI N18.7-1976. PGE has not compiled such a sununary document that indexes procedures and instructions to this Revision 3 1
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O O O AIR CO_ OLE _R ENC _LOSURE p_ _q ^4 @ l I l f^-- { AIR COOLERS h OUTSIDE AIR l l EXHAUST TO OUTSIDE M +^~ L _ _._I 99 99 = RADIATION DETECTOR _0 v h-{- 5 f 3 _r-$ N l m dC V l c FILTER DEMINERALIZER O_ Y y ~ SPENT FUEL POOL PUMPS l Figure 3.3-4 Modular Spent Fuel Pool Cooling and Cleanup System i Revision 7
4.0 OPERATIONS ,9 i j Q) This chapter discusses facility operations reated to the safe storage of irradiated fuel; including criticality prevention, chemistry control, instrumentation, maintenance activities, administrative controls of systems, spent fuel handling, and spent fuel cooling. To focus proper attention on important equipment and activities during the defueled condition, certain structures, systems, components, and activities have been categorized as important to the safe storage ofirradiated fuel. The following structures, systems, and components are categorized as important to the safe storage of irradiated fuel: (1) Safety-Related structures (includes the spent fuel racks; Spent Fuel Pool (SFP); fuel transfer canal; cask loading pit; fuel transfer tube; and the Control-Auxiliary-(v) Fuel Building complex). (2) Quality-Related installed monitoring equipment used to satisfy Trojan Technical Specification (TTS) surveillance requirements (includes SFP level monitoring and SFP temperature monitoring). (3) The Modular SFP Cooling and Cleanup System that provides the primary method l of SFP cooling and SFP water purification. l I (4) Portions of the Service Water, Domestic (Potable) Water, Fire Water Systems, and l portable gasoline-powered pump (s) taking suction from the Columbia River that l may provide makeup water to the SFP. l s
Activities that are important to the safe storage of irradiated fuel are controlled through use of appropriate approved procedures. O i Revision 7 4.0-2
4.1 OPERATION DESCRINION O 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 the Modular SFP Cooling and Cleanup System. In the event forced SFP cooling is lost, the l 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 CRITICAI.ITY 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 contained in Section 3.2.2. The design of the SFP is such that criticality prention is provided by maintaining center-to-center distance between adjacent fuel assemblies and the fixed neutron absorber contained in the spent fuel storage racks. The design of $e spent fuel storage rack assembly was such that it was impossible to insert spent fuel assemblies in other than prescribed locations, thereby maintaining the required separation and preventing accidental criticality. One storage rack has been removed from the SFP floor prior to the start ofISFSI loading. The removal of one SFP rack makes it possible to insert a fuel assembly into the open area that remains. If a fuel assembly is placed immediately along side a remaining rack, the assembly could be at a - spacing less than the minimum center-to<: enter distance required to ensure against accidental criticality. This condition was evaluated during the SFP rerack that installed these racks. The - prescribed compensatory measure was to add administrative controls to the fuel handling procedures that restrict storage of fuel assemblies in the cells of the remaining racks immediately adjacent to the opening left by removal of the one rack. This ensures that a fuel-assembly placed in the open space will not_be adjacent to another fuel assembly. The design assumes that all storage locations contain fuel assemblies with a maximum enrichment of 4.5 percent (U-235). Credit is not taken for boration of the SFP cooling water for bulk i 4.1-1 Revision 7 T r-"'"'-NF Y-T- v 9-'- 7
~ 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 transferred 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 l are employed. Plant operating procedures direct SFP temperature to be maintained between 40 F and 140 F and SFP boron concentration to be >_2000 ppm. Procedures also limit movement of loads over the spent fuel that could result in impact 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 ruptured as analyzed in Section 6.2.4. To minimize the probability l of a fuel handling accident, fuel handling operations are performed in accordance with approved Plant procedures under the direct supervision of a Certified Fuel Handler (CFH). 41.2 CHEMISTRY CONTROL l To minimize corrosion and contamination, chemistry of the SFP is maintained. Table 4.3-1 provides a listing of chemistry parameters monitored and their frequency. Revision 7 4.1-2
Corrective measures for out of specification chemistry will be taken upon discovery. The Modular SFP Cooling and Cleanup System is used to maintain satisfactory SFP purity and l \\ clarity. The capability for boric acid addition is required to ensure SFP coolant can be i maintained.2.2000 ppm.. Boric acid normally wal be mixed in the boric acid batch tank and 1 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 i SFP level, pump discharge pressure, temperatures throughout the system, and radiation level l from the air cooler enclosure drain line. Alarms are provided to the control room for l I abnormal SFP level or high SFP temperature. The radiation detector monitors for leakage l from the air coolers or piping inside the enclosure and automatically shuts down the system if l 4 I leakage is detected. There are no other automatic actions performed by SFP cooling l j, instrumentation with system operation being manually controlled. A more complete description of the SFP Cooling System is provided in Section 4.3. v l 4.1.3.1 Seismic Monitoring Instrumentation i j. The seismic monitoring instrumentatisn, consisting of a multi-element seismoscope and peak acceleration recorders, is subject to routine maintenance and testing to ensure it is available to record data from seismic events. SR 6341 Multi-element Seismoscope ' SR-6340D Peak Acceleration Recorder - Control Building Mezzanine, Top of Ladder Above Secondary Sample Storage Rooni SR-6340E Peak Acceleration Recorder - Fuel Building,93-foot Elevation Hot Shop, West, Behind the Wall O 4.1 3 Revision 7
SR-6340F Peak Acceleration Recorder - CCW Heat Exchangers,45-Foot Elevation Area 3 Base In the case where an instrument is removed from service for greater than 30 days, alternate means shall be provided to record the data needed for comparison with the design bases of facility features important to safety, unless it is determined by Engineering that the remaining instruments can provide sufficient data for analysis. Seismic monitoring instrumentation shall be in service during spent fuel cask loading operations. While it remains installed, maintained and tested, the time-history recording system provides an acceptable alternate means to record the data needed from a seismic event. ST-6336C Accelerograph - Fuel Building, 93-Foot Elevation ST-6336D Accelerograph - Cable Spreading Room ST-6336E Accelerograph - Free Field SR-6336A Seismic Recorder Unit h SR-6336B -Seismic Recorder Unit SR-6337 Seismic Playback Unit SS-6336 Seismic Trigger Within 24 hours following a seismic event data shall be retrieved from actuated instruments. The data shall be analyzed to determine the magnitude of the vibratory ground motion. A report shall be prepared and submitted to the NRC within ten days describing the magnitude, frequency spectrum, i.nd resultant effect upon facility features important to safety. 4.1.4 MAINTENANCE ACTIVITIES Maintenance activities include corrective and preventive maintenance as well as periodic l testing. Maintenance activities focus on the SFP Cooling System and components that support the emergency plan, fire protection plan, security plan or other licensed condition. Revision 7 4.1-4
4.3 SPENT FUEL COOLING AND SUPPORT SYSTEMS O 4.3.1 SPENT FUEL POOL COOLING The Modular SFP Cooling and Cleanup System is a closed-loop system. It takes suction from l the SFP to one of the two cooling pumps installed in parallel. The cooling pump discharge l 4 I - may be aligned through the purification filter and/or the demineralizer and back to the' pool. l When using the modular system, one or both of the air coolers may be operated continuously [ or on an intermittent basis to maintain desired SFP temperature. Even when not operating the l air coolers, flow may be maintained through the modular filter and/or demineralizer. The l I flowrate of one cooling pump is sufficient to recirculate a volume of water equal to the SFP l inventory through the filter and demineralizer approximately once every two days. This is l 4 adequate for removing fission products and other contaminants that may be introduced if a fuel l assembly were to leak. The cooling pump also provides skimmer flow. A small percentage of l the suction flow is drawn from the surface of the pool via the floating pool skimmer. This l portionjoins the main suction flow path. It may be aligned to pass through the purification l filter and/or the demineralizer, and/or the air cooler (s) as applicable. It is then discharged l l back to the pool. Figure 3.3-4 provides a diagram for the Modular SFP Cooling and Cleanup l j System. A discussion of the design bases is contained in Section 3.3.2. l Instrumentation is provided to monitor parameters necessary to demonstrate proper operation of the Modular SFP Cooling and Cleanup System. SFP temperature is periodically monitored l by control room indication and a high temperature alarm is provided. An alarm for high/ low SFP level is also provided in the control room. SFP level indication is p.ovided locally and checked periodically. Although SFP temperature and level are the only parameters required to ensure SFP cooling is being properly maintained, additional system instrumentation has been provided. SFP cooling pumps have discharge pressure gauges to monitor performance. l Temperature indication is located at various points in the system to allow for system testing. l and flow indication of SFP coolant upstream of the air coolers is also provided. l 4.3-1 Revision 7 n ,a-n
l A radiation detector monitors a drain line loop seal from the bottom of the air cooler l enclosure. This will detect leakage from a cooling coil tube or piping inside the enclosure. If l leakage is detected, the radiation detector automatically shuts down the cooling water pump l and cooling fans to stop a potential release of radioactive pool water to the environment. l In addition to the modular SFP cooling instrumentation described above, leak detection of the SFP liner plate is provided by trenches that collect leakage through the liner. The trenches l drain through normally open valves into the Liquid Radioactive Waste System. The leak detection valves are arranged into manifolds and the approximate location and magnitude of a leak can be determined by monitoring the leakage through these valves. These valves are checked periodically by operations personnel. I Chemistry of the SFP coolant is maintained to minimize corrosion and monitor for any contaminants that may be introduced. Table 4.3-1 lists the chemistry specifications and sampling requirements. 4.3.1.1 Off-Normal Operation of the Spent Fuel Cooling System Off-normal conditions for the spent fuel system can be categorized as loss of SFP level, loss of spent fuel cooling, and high SFP pool level. A brief description of the mitigating actions for each is provided below. 4.3.1.2 Loss of Spent Fuel Pool Level In the event of decreasing SFP level, Plant procedures direct the establishment of makeup l water to the SFP. Various water sources are available for makeup to the SFP. Among the j systems that can provide makeup to the SFP are: service water, domestic (potable) water, and l portable gasoline-powered pump (s) taking suction from the Columbia River. As a backup to the above water sources, Plant procedures provid: for use of fire main water as an emergency Revision 7 4.3-2
makeup water source. Systematic guidance is provided by Plant procedures for leak l identification and isolation. The only requirement to assure adequate cooling for the spent fuel is to maintain the water level in the SFP so that the spent fuel elements are not exposed. All lines entering the SFP which could siphon the pool to Elevation 76 feet 7 inches or below (equivalent to approximately 10 feet above stored fuel assembly), are equipped with siphon breakers to limit l SFP water loss to an elevation of 83 feet 11 inches. The SFP gates for access to the cask loading pit and refueling canal are equipped with inflatable seals to prevent level loss to these adjacent areas. For a postulated event wherein the SFP is drained to the Siphon Breaker level 1 followed by failure of the SFP gates and subsequent spillage of SFP water to the fuel transfer canal and cask loading pit (initially assumed empty), SFP level would reach a minimum level t of 76 feet 7 inches, or 10 feet above the top of the fuel assemblies. Additional discussion of the SFP design to prevent loss of level is contained in Section 3.2.2. Accident analyses for l loss of SFP level are contained in Section 6.3. lO 4.3.1.3 Loss of Soent Fuel Pool Cooling l As discussed in Section 4.3.1, a high temperature alarm is provided in the control room with a maximum setpoint of 135*F. Plant procedures direct actions to restore operation of SFP j cooling. As stated in Section 4.3.1.2, the only requirement to assure adequate cooling for the j spent fuel is to maintain the water level in the SFP such that the spent fuel elements are not L exposed. Procedure guidance directs the restoration of level by the use of makeup to the pool l l in the event oflevel loss from boil off. Restoration of SFP level is discussed in Section 4.3.1.2, Loss of Spent Fuel Pool Level. l b 4.3-3 Revision 7 l l
4.3.1.4 High Snent Fuel Pool Level A level switch has been provided in the SFP for the purpose of transmitting high and low water level annunciation signals to the control room. Plant procedures are in place that provide guidance for the systematic determination and isolation of the water source. 0"erflow from the SFP is directed to the dirty waste drain tank. 4.3.1.5 Safety Criteria and Assurance The only requirement to assure adequate cooling for the spent fuel is to maintain the water level in the SFP so that the spent fuel elements are not exposed. The top of fuel seated in the spent fuel storage modules is located at Elevation 66 feet 7 inches. Dose calculations for fuel handling accidents assume a minimum of 23 feet of water above the top of spent fuel assemblies; therefore, SFP level less than Elevation 89 feet 7 inches requires operator action to restore level. This minimum required level also ensures dose rates at the SFP surface will be maintained <2.5 mrem /hr during fuel movements. A SFP minimum boron concentration of 2000 ppm is assumed for certain design basis accidents to prevent inadvertent criticality. Sampling frequency and corrective actions to ensure boron concentration is maintained have been implemented. A maximum temperature limit of 140 F was assumed for SFP bulk temperature. This limit ensures that for the worst case loss of SFP cooling accident, SFP boiling will not occur within 49 hours, and that at least 10 days are available to establish makeup to the SFP before boil off would reduce SFP level to 5 feet above the fuel assemblies. Maintaining 5 feet of water above the active portion of the fuel provides adequate radiation shielding to allow access for restoration of SFP level. Table 6.3-2 provides estimates of radiation dose rates for various SFP levels. Revision 6 4.3-4
.m 1 4.3.2 EI ECTRICAL DISTRIBUTION l During normal Plant operation, Plant electrical demands are provided from the 230-kV j switchyard. The design and reliability of the offsite power source are discussed in Section 3.4.1. Offsite power is normally supplied to the Plant distribution system via the startup transformers. i 1 In the event of loss of offsite power, forced cooling capability would be lost. Without l restoration of forced spent fuel cooling, boiling of the SFP would not occur for at least 88 hours. SFP inventory is adequate to allow boiling for a minimum of 22.5 days without any source of makeup and still maintain > 10 feet oflevel above the top of the spent fuel ' assemblies. Makeup capability to the Spent Fuel Pool independent of Plant power sources can l . be provided by the fire main system (diesel-driven fire pump). A discussion of the fire main system is contained in PGE-1012, " Trojan Nuclear Plant Fire Protection Program." I o I { 4.3-5 Revision 7 l c =-- - - - - -='-r
4.4 CONTROL ROOM AREA The Control Room provides a centralized station to monitor SFP temperature and level. At l least one person qualified to stand watch in the Control Room shall be present in the Control Room when irradiated fuel is stored in the SFP. A discussion of the activities normally performed by the Control Room operator is provided below. SFP instrumentation available in the Control Room includes: SFP temperature indication, SFP l high temperature alarm, and SFP level alarm (high/ low). In the event Control Room instrumentation becomes unavailable, the Control Room operator can utilize local monitoring of spent fuel pool temperature and level. All operation of the SFP Cooling System, including normal and emergency makeup, is performed locally since remote operation of SFP ccmponents and makeup flow paths is not provided in the Control Room. i Operatio.1 of the Plant's electrical distribution system is normally controlled and monitored in the Control Room. Local operation of the electrical distribution switch gear and load breakers can be performed if required. l Additional systems that may be monitored or controlled in the Control Room include: (1) Plant Ventilation (2) Area Radiation Monitors (3) Process Radiation Monitors (4) Fire Protection 4.4-1 Revision 7
5.2.5 MODULAR SPENT FUEL POOL COOLING SYSTEM COOLING AIR l O i The Modular SFP Cooling System uses outside cooling air to cool the SFP. This system is not l considered a building ventilation system since it provides no building habitability functions. l l The system design bases are described in Section 3.3.2. l l As noted in Section 5.2.3, the Fuel and Auxiliary Building Ventilation System maintains a l negative atmosphere in spaces subject to airborne radioactive contamination. When both l Modular SFP Cooling System cooling fans are running, the portion of the air cooler enclosure l upstream of the fans operates at a slight negative pressure relative to the Fuel Building. This l is acceptable since the enclosure is separate from the Fuel Building environment. There is no l mixing of the outside cooling air with the air inside the Fuel Building and the enclosure does j not provide a release path for airborne radioactive material that might be present inside the l Fuel Building adjacent to the enclosure. l O l 5.2-9 Revision 7
5.3 LIOUID WASTE TREATMENT AND RETENTION Ov The Liquid Waste Treatment and Retention System (LRWS), shown in Figure 5.3-1, typically l collects, stores, processes, monitors, and discharges plant liquid effluents that may contain l radioactive nuclides. The system consists of compenents from the fonner Clean Radioactive - l Waste Treatment System (CRWS) and the Dirty Radioactive Waste Treatment System (DRWS). l The CRWS and DRWS have been consolidated by plant modification to form the LRWS. Where - l other licensing documents refer specifically to the CRWS or DRWS, these terms are now l interchangeable with the LRWS. l l 13.1 Design Bases l l The following design bases apply to the LRWS: l 1 (1) To provide a means for collecting liquid effluent from floor and equipment drains and l sumps in the Containment Building, the Fuel Building, and the Auxiliary Building l l (2) To provide sufficient storage capacity for the maximum anticipated liquid flow from l supported systems l l (3) To reduce the concentration of nuclides and particulates in the stored liquid to a level l that will permit release to unrestricted areas within the guidelines of Appendix I to l ) 10 CFR 50, and limit the dose to any organ to < 5 mR during any calendar yearm l 1 (4) To provide storage capability in the discharge path for monitoring the liquid prior to l release and for recycling those batches that do not meet the design objectives l l (5) To measure and record the release of plant liquid radioactive effluent l l O 5.3-1 Revision 7 i i
l 5.3.2 System Descriotion l l The typical sources of liquids collected in the LRWS are various floor and equipment drains, l washdown water from decontamination evolutions, and flush water. Processed effluent from the l LRWS is discharged. l l The system consists of the following components: I l (1) Two Treated Waste Monitor Tanks (TWMT) l l (2) Two Treated Waste Monitor Tank pumps l l (3) Dirty Waste Drain Tank (DWDT) l l (4) Two Dirty Waste Drain Tank pumps (one electrically deactivated) i O l (5) Auxiliary Building sumps I l (6) Containment Building sumps I l (7) Two influent bag filters l l (8) Two effluent bag filters l l (9) Liquid radwaste discharge process and effluent radiation monitor (PRM-9) I l (10) Interconnecting piping, valves, and instrumentation l l Liquids processed by the LRWS may contain varying amounts of boric acid and other chemicals. O Revision 7 5.3-2 1
Piping, valves, and major components (with the exception of certain radwaste pump strainers) in l contact with process fluids are constructed of corrosion resistant materials. -l 1 System tanks are maintained at near-atmospheric pressure and are vented to the vent collection l header. Tank overpressure protection is provided by tank overflows to the plant floor drain l . system. l 1 - Containment and Auxiliary Building sumps collect drainage from various plant floor and l equipment drains. The discharge from the various sumps is routed to the DWDT. l l The DWDT is divided into two equal sections. The DWDT pump can be aligned to either tank l section. The pump is typically used to transfer tank contents through the influent bag filters to l the treated waste monitor tanks. -l l The influent bag filters are normally used to remove particulate matter discharged from the l r system collection tanks. Differential pressure indication is provided to alert the operator to l excessive filter clogging that will require filter replacement. l I The TWMTs receive fluids from the system collection tanks and store the fluids for monitoring l and possible further processing prior to discharge. The TWMT pumps can be aligned to either l tank. The pumps are typically used to recirculate the tank contents for sampling prior to l discharge, transfer the waste from tank to tank through the effluent bag filters (and/or optional l demineralizer) for additional processing, or to discharge the tank contents. Mixing eductors on l recirculation lines, internal to the tanks, are designed to achieve thorough mixing to ensure l representative sampling. l l The effluent bag filters are normally used to remove particulate matter from the TWMT contents, l if additional processing is determined to be required prior to discharge based upon sample results. l Differential pressure indication is provided to alert the operator to excessive filter clogging that l 5.3-3 Revision 7
l will require filter replacement. Additionally, connections exist for an optional demineralizer if l filtration is not adequate to support discharge requirements. l l The plant discharge header receives liquid waste from the TWMTs and discharges the waste to ! the Columbia River by way of the discharge and dilution structure. The plant discharge header l flow signal is used to regulate plant discharge flow by controlling the header flow control valve, l ' which maintains the discharge rate at a preset value. PRM-9 monitors the discharge header for l radioactivity and actuates the header isolation valve to stop discharge flow if excessive l radioactivity is detected. Section 5.5 provides further discussion of PRM operation. In addition, l discharge flow will be stopped whenever insufGcient dilution flow is detected at the discharge and l dilution structure. Flush water for header cleanup following a discharge may be provided from l various non-contaminated sources. l l The discharge and dilution structure is discussed in Section 3.3.9. l l 5.5.3 Design Evaluation l l The LRWS does not perform any safety functions for the permanently defueled plant condition. l The system collects waste water from various sources and processes it for discharge. The LRWS l l has adequate capacity to process waste from the small number ofinput sources in operation with l the plant in the permanently defueled condition. l l Administrative controls are established to require further processing when necessary to ensure l effluent releases are within allowable limits. l l O l Revision 7 5.3-1 l l I
....._~...._ _._ _ _ 5.4 SOLID WASTES The Solid Radioactive Waste System (SRWS) provides for storage and processing for disposal of ' radioactive solid wastes generated at the Plant. These wastes include spent demineralizer resin, expended filters, and miscellaneous contaminated equipment and solid refuse. l l - 5.4.1 DESIGN BASES l The design bases of the SRWS include the following: I (1) To provide a means of remotely removing and transferring contaminated expended l filters from the filter vessel to the solid radioactive waste container selected for the filter in a manner which minimizes exposure to operating personnel (2) To provide a means of packaging spent demineralizer resins and expended filters in l disposable containers suitable for transfer from the Plant site L (3) To provide adequate shielding in storage areas for retaining wastes in di.sposable l containers pending shipment to appropriate disposal facilities (4) To provide a means of packaging miscellaneous contaminated solid wastes generated by l sampling, ventilation filter replacement, decontamination refuse and various other items L resulting from maintenance (5) To utilize shipping contsiners and procedures which conform with the regulations of l L Title 10, Parts 20 and 71, and Title 49 of the Code of Federal Regulations l l (6) To observe, measure, and record the contamination levels of solid radioactive wastes l which are processed for shipment from the site l 5.4-1 Revision 7 l. ~
) 5.4.2 SYSTEM DESCRIPTION O Inputs to the SRWS include the following: (1) Spent demineralizer resins which have been used to process potentially radioactive liquid (2) Expended filters which have been used to process potentially radioactive liquid (3) Miscellaneous solid wastes which are potentially contaminated l 5.4.2.1 Spent Resin Transfer System l l Spent demineralizer resins are transferred directly to a disposable liner. Water is removed from the disposable liner and returned to the Liquid Radioactive Waste System. s 5.4.2.2 Filter Handling In general, potentially contaminated filters will require replacement when filter clogging causes excessive differential pressure across the filter or when radiation levels from filter sludges become excessive. 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. O Revision 7 5.4-2
I: 5.4.2.3 Solid Wastes O Dry, solid radwaste processing is typically performed in the Radwaste Annex, Containment, free release facility, or in the Radwaste. Processing Building (formerly the Condensate Demineralizer Building). Some segregation of radwaste material takes place within the plant buildings. The
- Radwaste Processing Building is typically used for cutting, decontamination, and/or packaging solid radioactive waste.
The Radwaste Annex to the Fuel Building includes a drum compactor, which is used for compacting dry, active wastes. A free release facility (formerly the Emergency Diesel Generator rooms) is used to perform final surveys of decontaminated solid radwaste intended for free release. A solid waste compactor may be used to compact miscellaneous solid waste materials - into drums for storage'and shipment offsite. 5.4.3 DESIGN EVALUATION O The SRWS does not perform any safety functions. The volume of spent resin required to be processed with the Plant permanently defueled is significantly less than with the Plant operating. l In addition, the maximum expected activity of the spent resin volume is expected to be far less than with the Plant operating. The maximum expected activity of the spent resin volume was conservatively based on the resin fission product activities for Plant operation with reactor coolant activity levels determined on the basis of fission product diffusion through cladding defects in 1.0 percent of the fuel rods. Similarly, volume and maximum expected activity associated with expended filters and miscellaneous solid wastes for the permanently defueled Plant condition are bounded by the analysis for Plant operation. b f l O i V I' 5.4-3 Revision 7 l ,w y -<,-.y-,,.. w-c.,, w -v- -+-m-m.-+- -~ ---,- - - - +
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5.5 PROCESS AND EFFLUENT MONITORING SYSTEMS V The process and effluent monitoring systems (PERMS) provide the means for monitoring the . gaseous and liquid effluent flow paths by which radionuclides may be released to the environment. The sysams consist of permanently installed monitoring devices with a program for sample collections and data analyses. This section describes the instrumentation and the program for sample collection and analysis at Trojan. 5.5.1 DESIGN BASES The principal objectives of the PERMS are: l (1) To evaluate the performance of Plant systems that minimize the release of radioactivity O v to accessible areas of the Plant and to the environment. (2) To estimate quantities of radioactivity in effluents or potential effluents before and/or -during releases to unrestricted areas. (3) To provide a method to quantify releases of radionuclides. l Specific design objectives include the following: i (1) To collect representative samples of gaseous and liquid effluents to unrestricted areas to allow measurement of the principal radionuclides present in these discharges as required j by 10 CFR 50. 5.5-1 t 1 i
(2) To monitor radioactivity in both intermittent and continuous discharges of potentially radioactive gaseous and liquid effluents to unrestricted areas to verify that the concentrations and radiation levels are within the limits of 10 CFR 20 and 10 CFR 50, Appendix I. (3) To provide alarm and automatic termination of the release of effluents when radionuclide concentrations and radiation levels exceed the limits specified above in Paragraph (2). Where the termination of the release is not feasible, the monitors provide indicators of the magnitude of the activity released. 5_.5.2 SYSTEM DESCRIPTION The liquid and gas radiation monitoring systems are presented in detail in the following subsections. The effluent monitor setpoints and their bases are presented in the Trojan Nuclear Plant Offsite Dose Calculation Manualm. O Indicators and controls for liquid and gaseous process and effluent monitoring systems are provided in the Control Room at the radiation monitoring panel. Controls are also provided at l the individual local monitor instruments. Exceptions are made to the Liquid Radioactive Waste l Discharge Monitoring System (PRM 9) where indicators and controls are provided in a local l l ratemeter but not in the Control Room radiation monitoring panel. Multi-point recorders are provided to continuously record the detector outputs from the monitors described in the following [ sections. Each ratemeter indicator is provided with a triple-level alarm system. One of the alarms indicates high voltage or power failures. The other two alarms serve as sequential alarming indicators of increasing radiation levels. The Control Room radiation monitoring panel l provides a common alarm indicating panel for these alarms, except PRM 9 where alarms are l annunciated in a Control Room PC station. l Revision 7 5.5-2
Each gaseous or liquid monitoring system is provided with a sample chamber that is sized and shielded as required to achieve the required sensitivities. Each sampler is constructed to L instrument design standards consistent with their service and the design of the process system to ? . which it is connected. Samplers are located as close as practical to the process stream such that sample line interference will be insignificant. 5.5.2.1 I iauld Monitoring Systems .t 5.5.2.1.1 Liquid Radioactive Waste Discharge Monitoring System (PRM-9) PRM-9 monitors the activity of liquids discharged to the Plant liquid radioactive waste discharge header. The header receives liquid wastes from Plant systems that potentially contain radioactive i nuclides and constitutes the path by which planned releases occur from the LRWS. Section 5.3 l provides a description of the Liquid Radwaste System. l The monitoring system consists of a NaI gamma scintillation detector mounted externally on the [ process piping. The radiation levelr. are displayed on a log ratemeter at the Plant liquid l ) radioactive waste discharge header. The ratemeter provides alert and high alarms which are - l annunciated in a Control Room PC station. The high alarm setpoint is established in accordance. l with the Trojan Nuclear Plant Offsite Dose Calculation Manual *. Upon receipt of a high alarm . or circuit failure signal, the Plant discharge header automatic isolation valve will close to l immediately discontinue discharge flow to the discharge and dilution structure. The system L response and isolation valve location are such that high activity liquids will not be released before the valve closes. ~ 5.5.2.1.2 Liquid Sample Collection System j . The Liquid Sample Collection System is provided to allow sampling at the intake structure and at the discharge and dilution stmeture for evaluation of the adequacy of :nplant monitoring. 5.5-3 P.evision 7 p + - -.. ~.
The sampler at the discharge and dilution structure consists of a head tank which maintains constant level above the solenoid valve. The solenoid valve is provided with a timer to allow for automatic sampling at adjustable frequencies and duration. The total sample volume collected can be manually adjusted by changing the time interval between resetting of the sampling timer and by adjusting the outlet throttle valve. 5.5.2.2 Gas Monitoring Systems The following gas monitoring systems are provided: (1) Containment Monitoring System (PRM-1) (2) Auxiliary Building Vent Exhaust Monitoring System (PRM-2) (3) SFP Vent Monitoring System (PRM-3) O (4) Vent Collection Header Monitoring System (PRM-5) Each of the gas monitoring systems consist of one or more detector channels dependent upon the service of the monitor. l l The airborne particulate channel consists of a filter sampler and a beta scintillation detector that is capable of detecting the radioactivity emitted from the accumulated particulates on the filter. The low-level and intermediate-level gas channels use a common shielded sample chamber containing a beta scintillation detector for low-level gaseous activity and a halogen-quenched Geiger-Mueller detector for intermediate-level activity. The two detector channels have one decade of overlap. Revision 7 5.5-4
I A three-way valve is provid:d at the sampler inlet to permit air purging of the sampler to facilitate background checks. The offline gas monitoring systems (PRM-1 and PRM-2) utilize sample pumps to draw the 1 necessary process gas samples through the sampler. The pumps provide a constant rate of l sample flow irrespective of changes in flow resistance through the filter media. Sample lines for ofGine samplets are fabricated of stain!ess steel. Sampling devices and procedures reflect the recommendations of ANSI 13.1-1969, Guide to Sampling Airborne Radioactive Material in Nuclear Facilities *. An evaluation of sample line losses of 1 radioiodine and air particulates was performed for PRM-1 and PRM-2. Line losses were found to be negligible. Each offline gas monitoring system is provided with a grab sample connection that can be used to obtain representative samples for laboratory analysis and with fixed-filter assemblies for collection of radioiodine, air particulates, and tritium. The following gas monitoring m:thod(s) are also provided: (1) Radwaste Processing Building monitoring method ) The gas monitoring method utilized for the Radwaste Processing Building consists of an airborne particulate filter. A, sample pump is used to draw the necessary gas sample through the sampler. l 5.5.2.2.1 Containment Monitoring System (PRM-1) i PRM-1 monitors the air part culate activity levels in %e purge exhaust duct when purging. i The monitor quantitatively analyzes airborne activity released to the environt by the t Containment Purge Exhaust System. Section 5.2.2 provides a description of the Containment F i i Purge Exhaust System. 5.5-5 Revision 7 l 4
The Containment Monitoring System utilizes the air particulate channel (PRM-1 A). PRM-1 utilizes a single sample pump that draws a single gas stream through the airborne particulate monitoring moving filter. During containment purge operations, a sample is drawn from and returned to the purge exhaust duct by an isokinetic sampling probe. The system may also be aligned to sample the containment atmosphere. 5.5.2.2.2 Auxiliary Building Vent Exhaust Monitoring System (PRM-2) PRM-2 monitors the gt.seous and air paniculate activity levels released to the environment by the combined ventilation exhaust flows from the Fuel and Auxiliary Buildings. The system also monitors releases from the SFP ventilation exhaust and the vent collection header after they are diluted by the Fuel and Auxiliary Building ventilation exhaust flow. Section 5.2.3 provides a description of the Fuel and Auxiliary Building Exhaust System. The Auxiliary Building vent Monitoring System utilizes the following channels: (1) Air particulate (PRM-2A) (2) Low-range gas (PRM-2C) (3) Intermediate-range gas (PRM-2D) PRM-2 utilizes a single sample pump that draws a single gas stream in series through the airborne particulate monitoring moving filter and a low-and intermediate-level gas monitoring chamber. Sample flow is drawn from and r: turned to the purge exhaust duct by an isokinetic sampling probe. Revision 7 5.5-6
y--, t -- gg--- g 5 i. _. I.-% TABLE 5.5-1 -I RADIOLOGICAL ANALYSIS
SUMMARY
OF LIQUID PROCESS SAMPLES Analytical Sampline Location Frequency Analysis Performed ~ Sensitivity (uCi/cci - Purnose of Samnie Service Water System l. ' At intake from Columbia River Weekly composite Gross beta 5 x 10-8 Background determination Tritium 10 ~ 4 7 Liquid Radwaste System I Treated waste monitor tanks Each batch discharged Gamma spectrometric 5 x 10-' Determine whether to process or discharge ' ? Modular SFP Cleanup System ],' SFP water Monthly Gamma spectrometric 5 x 10-' Determine purification requirements Downstream of demineralizer Monthly Gamma spectrometric 10 Evaluate performance ofion exchangers 5 Main outfall of Discharge Weekly composite Gross Beta 5 x 10 ' Indication of Service Water System leakage 8 ard Dilution Structure Tritium 104 Backup monitoring for releases; verification of conformance with 10 CFR 20 4 Revision 7 b -.a-r ~
. _... _ _ _. _ _ -... ~... - _ - - - _ _ - - - - - - - - -. - - - - - - - - - - - ~ - - - _ - - - - _ - - - _ - - -. _ - - - - -. - - - - - - - - - - - - - _ - - - - - - - - - -. - -. - _ - _ _ O Oi OI l 1 i t .t HEPA AB-4 DAMPERS. FILTERS fillers -DAMPERS 4 8. b Z W Boo. .t i C-TRAlli N t 6 1 1 1 A-TPAlli FROM FUEi. AllD u e AU ~tLIARi EUILD!!!GS L 2 I 0- TRAlti N PERM [ 2 i 'l- - i r N-m B 1 U B00 1 i r u B-IRAlti 10 [XHAUSI' l I k i Figure 5.2-5 Fuel / Auxiliary Building Ventilation Exhaust System (AB-3) j l
4 'i l !4, I i. F v U v ~ O w 0 n n = H E A T ER e-D AM / \\ P l E l R S F igu re 5 = = 2 F I 6 L T V E e R n S t i la e t ion Ex A = = B h p a H g u E s T T P t R R A A g A S I I y N N s te m (AB D = = A l/ W X 4 ) ER S F AF O O A BR N - O S 3M 'Q 'k B DD S = P E R l M 3 = Re T v O e is E io X n H A 7 US T i 1ll1I 1lIl ll)4l llI)l)ll lIIll l l1Il l
O O O DISCHARGE & DILUTION STRUCTURE o I I RADIATION I MONITOR l---- tPRM-9) I -El-g stra--} g jNFLUENT FILTERS LJ 5 g#v-b t ,? ? INPUTS LJ d8 d8 EFFLUENT o 4 r8-- b FILTERS VN l i 8 ~ / N ^ 8 l i _ _ _ _i y DIRTY WASTE TREATED WAFTE OPil0NAL DRAIN TANK MONITOR TANKS DEMINERALIZER CONTAINMENT BLDG < i N / LJ N/ u u FLUSH t.UXILI ARY BLDG DWDT PUMPS TWMT PUMPS SUMPS WATER Figure 5.3-I Liquid Radioactive Waste System Revision 7
l 6.3 SPENT FUEL POOL ACCIDENTS
- O (L,.1 LOSS OF SPENT FUEL DECAY HEAT REMOVAL CAPABILITY 3
1 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 heat load associated with the 781 l spent fuel assemblies stored in the SFP, after five years following the 1993 permanent l shutdown and defueling of the Plant, (2) maintain fuel cladding integrity in the event all forced l 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 considered 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 allosvs adequate time (a minimum of ten days) to establish makeup capability to ensure that fuel elements remain covered. 6.3.1.1 Potential Events Resulting in Loss of Soent Fuel Decav Heat Removal Caoability 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: explosions, toxic chemicals, fires, ship collision with intake structure, oil or corrosive liquid spills in the river, and cooling tower collapse. Explosive hazards were analyzed for the Trojan site and det..inined not to result in failure of any safety-related equipment. The SFP is a safety-related structure however the SFP Cooling System is not. To be conservative, the explosive hazard was considered to be an initiating l 6.3-1 Revision 7 l 4 ... ~
event that could result in the damage of the SFP Cooling System such that a loss of inventory occurs. TI e consequences of this accident are discussed in Section 6.3.3. h Toxic gas bazards discussed in Section 2.3 constitute a hazard to personnel, not Plant 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 SFP Coeling System integrity therefore loss of inventory is not considered. The consequences of this s:c! dent are discussed in Section 6.3.2. l l Fire hazards were also considered. A fire is considered to render the SFP Cooling System inoperable. SFP Cooling System integrity is not considered to be damaged by the fire therefore loss of inventory is not considered. The consequences of this accident are discussed in Section 6.3.2. l l Oil or corrosive liquid spills in the river were also considered As discussed in f l Section 2.2.3.5, it is unlikely that this event would have any impact on the Trojan Facility. f Anticipated severe meteorological events have also been evaluated for the Trojan Facility. These events include tornadoes and high winds. 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 meteorological conditions were considered to result in loss of forced SFP cooling but not to impact SFP Cooling System integrity. Section 3.1.3.2.3 states that loss of SFP inventory due to a tornado is not anticipated to be of any appreciable amount, therefore inventory loss was not considered. Section 6.3.2 provides a discussion of the consequences associated with this event. l With the exception of off-site electrical distribution, the SFP cooling components are located above the maximum expected flood level. Loss of forced SFP cooling was considered credible Revision 7 6.3-2
due to the loss of electrical power. SFP Cooling System integrity is not affected therefore loss l of SFP inventory was not considered. Section 6.3.2 provides a discussion of the consequences associated with this event. Geological events were also evaluated. Hazards due te seismic activity and volcanic activity were evaluated. A discussion of these events is provided in Section 2.5. The loss of forced SFP cooling and loss of SFP inventory were considered to be possible results of these events. Section 6.3.3 provides a discussion of the consequences associated with this event. In addition to the above, system / component failures which could affect forced SFP cooling were also analyzed. Those systems whose failure could result in a loss of forced SFP cooling are: (1) Modular SFP Cooling System l (2) Loss of electrical power l l The loss of electrical power cannot impact SFP inventory. The consequences for this failure l i are bounded by the consequences presented in Section 6.3.2. l The worst case SFP system failure would be a loss of the original SFP Cooling System l piping / component integrity which results in reducing SFP to the level of the original system l piping siphon breakers. The initial condition for this type of failure is bounded by the l consequences presented in Section 6.3.3. Failure of the SFP structure which could result in reducing the SFP level below that assumed in Section 6.3.3 are not considered credible based on the design criteria presented in Section 3.2. i N% tU 6.3-3 Revision 7
6.3.2 LOSS OF FORCED SPENT FUEL COOLING WITHOUT CONCURRENT SFP INVENTORY LOSS The evaluation of a loss of forced SFP cooling without a concurrent loss of inventory was performed. To ensure that the consequences were bounding for this type of accident the following conservative assumptions were made: (1) SFP temperature was assumed to be 140 F at start of accident. This is the maximum temperature allowed by Plant procedures. (2) SFP level was assumed to be 23 feet above the fuel. This is considered conservative since normal level is maintained 24 feet above the fuel and Technical Specifications require the 23-foot minimum. Decay heat from the spent fuel at 40 weeks after shutdown is estimated using the methodology of ANSI /ANS Standard 5.1-1979 (Reference 1). The modeling of the spent fuel decay heat was based on the power history of the Plant. Individual rod histories were not used. The power operation for each month was converted to an equivalent operating time at full power which was assumed to occur at the end of each month. This is conservative in that it places more fission events near the end of each month and allows less time for decay to occur during a month. SFP decay heat generation for a decay time of 40 weeks after shutdown was utilized in the analysis. Figure 6.3-1 demonstrates the effects of decay time on the heat generated in the pool which increases the safety margin with increasing time since reactor shutdown. The evaluation showed that under the assumed conditions and at 3% years after shutdown SFP boiling would not occur for at least 88 hours. Boil off rate would be no greater than 3.7 gpm and at least 540 hours (22.5 days) would be available to establish a make-up source and still maintain at least 10 feet of water level over the fuel. Figure 6.3-2 provides a graph of SFP heatup rate versus time after reactor shutdown. This graph demonstrates that additional response time becomes available as time from reactor shutdown increases. Revision 7 6.3-4
Figure 6.3-3 provides a graph of time available prior to SFP boiling based on the time after l shutdown. The time to boil is based on an initial SFP temperature of 140*F. Figure 6.3-4 provides a graph of calculated boil off rate based on time after reactor shutdown. Figure 6.3-5 provides the makeup rate (based on a makeup water temperature of 100 degrees F) that would be needed to maintain SFP level constant until forced SFP cooling could be restored. Time available to establish a makeup source prior to SFP pool level being reduced to 10 feet above the fuel due to boil off (i.e. no inventory loss from system leakage) is provided in Figure 6.3-6. For inventory losses due to boil off only, more than 3 weeks are available to establish a makeup water source. l 6.3.3 LOSS OF FORCED SPENT FUEL COOLING WITH CONCURRENT SFP INVENTORY LOSS Certain events exist which could result in loss of inventory of the SFP. The events that could result in loss of inventory are discussed in Section 6.3.1.1. Regardless of the initiating event a bounding set of conditions can be established and the bounding consequences evaluated. Three events were identified in Section 6.3.1.1 that could result in a loss of inventory. An l explosion, a seismic event, and SFP Cooling System pipe failure (original and/or Modular l System) were considered to have the potential to affect the integrity of the SFP Cooling l System. The seismic event was considered to be the most bounding since this event had the l potential to affect all components associated with the original SFP Cooling System including l support systems and the new Modular System. Although the Modular SFP Cooling System l now provides forced cooling to the SFP, failure of the original SFP Cooling System, with its l potential to siphon the SFP to a lower level, has been retained as the bounding loss of l inventory event. The following bounding initial conditions were assumed to exist at the time l j of a seismic event: i (1) Initial SFP temperature was 140 F (2) The fuel transfer canal is empty with gate closed j 6.3-5 Revision 7 l
l (3) The fuel cask loading pit is empty with the gate closed Ol l The seismic event is assumed to cause a failure in the original SFP cooling System which results in the instantaneous draining of the SFP to the level of the siphon breakers. The gates l separating the fuel cask loading pit and fuel transfer canal then fail resulting in additional loss of level until equalized with the SFP level. This sequence of events, though unlikely, is conservative in that it maximizes SFP inventory loss. The resultant SFP level for this event l would be 76 feet 10 inches or 10 feet 3 inches above the fuel. The Modular SFP Cooling System and SFP structure design are discussed in Chapter 3. Based on this design criteria this event is considered to be the maximum credible inventory loss prior to boiloff. The remaining discussion will demonstrate that adequate time is available to establish a source of makeup water to the SFP such that uncovery of the fuel and loss of spent fuel cooling is not credible. Heatup rates were conservatively calculated assuming a starting water level of only 10 feet above the fuel. To account for the possibility that pool heat up may not be uniform additional conservatism was incorporated by reducing the available inventory by 10 percent. Credit was not taken for heat loss to the pool walls or for water available in the fuel cask loading pit or fuel transfer canal. The time for the SFP to boil was conservatively calculated to be 49 hours. The boil off rate would then be no greater than 3.7 gpm. At least 10 days are available to establish a makeup water source to the pool prior to a level reduction to 5 feet above the fuel. This calculation showed that personnel access for recovery actions was possible with the coolant level reduced to 5 feet above the fuel. Calculations were performed to determine maximum expected dose rates in the area above the l j SFP. Table 6.3-2 provides the results of these analysis. Recovery of SFP level and cooling are discussed in Section 4.3.1. O Revision 7 6.3-6 l l ,}}