LR-N17-0088, Redacted - Hope Creek Generating Station Updated Final Safety Analysis Report, Rev. 22, Chapter 12, Radiation Protection
Text
SECTION 12 RADIATION PROTECTION TABLE OF CONTENTS
Section Title Page
12.1 ENSURING THAT OCCUPATIONAL RADIATION 12.1-1 EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE 12.1.1 Policy Considerations 12.1-1
12.1.1.1 Management Policy 12.1-1 12.1.1.2 Management Responsibilities 12.1-1 12.1.1.3 Policy Implementation 12.1-4 12.1.2 Design Considerations 12.1-5
12.1.2.1 General Design Considerations 12.1-6 12.1.2.2 Equipment Design Considerations 12.1-7 12.1.2.3 Facility Design Considerations 12.1-8 12.1.2.4 ALARA Model Design Review Program 12.1-10 12.1.2.5 Field ALARA Design Program 12.1-13 12.1.3 Operational Considerations 12.1-14
12.1.3.1 Procedure Development 12.1-15 12.1.3.2 Station Organization 12.1-16 12.1.3.3 Operating Experience 12.1-16 12.1.3.4 Exposure Reduction 12.1-17 12.1.3.5 General ALARA Techniques 12.1-17 12.1.4 References 12.1-20
12.2 RADIATION SOURCES 12.2-1 12.2.1 Contained Sources 12.2-1
12.2.1.1 Primary Containment (Drywell) 12.2-2 12.2.1.2 Reactor Building 12.2-5 12.2.1.3 Service and Radwaste Areas of the 12.2-9 Auxiliary Building
12.2.1.4 Turbine Building 12.2-11
12-i HCGS-UFSAR Revision 0 April 11, 1988 TABLE OF CONTENTS (Cont)
Section Title Page 12.2.1.5 Shielding Design Sources Resulting from 12.2-13 Design Basis Accidents
12.2.1.6 Stored Radioactivity 12.2-13 12.2.1.7 Special Sources 12.2-14 12.2.2 Airborne Radioactive Material Sources 12.2-15
12.2.2.1 Sources of Airborne Radioactivity 12.2-15 12.2.2.2 Emission of Airborne Radioactive Materials 12.2-15 12.2.2.3 Locations of Sources of Airborne 12.2-15 Radioactivity
12.2.2.4 Control of Airborne Radioactivity 12.2-15 12.2.2.5 Methodology for Estimating the 12.2-16 Concentration of Airborne Radioactive Material within the Plant 12.2.3 References 12.2-19
12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 Facility Design Features 12.3-1
12.3.1.1 Common Equipment and Component Layout 12.3-1 and Designs for ALARA
12.3.1.2 Common Facility and Layout Designs for 12.3-9 ALARA 12.3.1.3 Radiation Zoning and Access Control 12.3-15 12.3.1.4 Control of Activated Corrosion Products 12.3-16 12.3.2 Shielding 12.3-17
12.3.2.1 Design Objectives 12.3-17 12.3.2.2 General Radiation Shielding Design 12.3-19 12.3.2.3 Shielding Calculational Methods and 12.3-33 Geometry Models 12.3.3 Ventilation 12.3-35
12.3.3.1 Design Objectives 12.3-35 12.3.3.2 Design Criteria 12.3-35 12.3.3.3 Design Guidelines 12.3-36
12-ii HCGS-UFSAR Revision 16 May 15, 2008 TABLE OF CONTENTS (Cont) Section Title Page 12.3.3.4 Design Description 12.3-41 12.3.4 Area Radiation and Airborne Radioactive 12.3-42 Materials Monitoring Instrumentation 12.3.4.1 Area Radiation Monitoring Design 12.3-42 12.3.4.2 Airborne Radioactive Materials 12.3-49 Monitoring 12.3.5 References 12.3-52 12.4 DOSE ASSESSMENT 12.4-1 12.4.1 Direct Radiation Dose Estimates for 12.4-1 Exposures Within the Plant Structures 12.4.1.1 Definition of Categories Used in 12.4-2 Exposure Estimates 12.4.1.2 Exposure Estimate Methodology 12.4-3 12.4.1.3 Results of Annual Direct Radiation Dose 12.4-5 Estimates 12.4.2 Airborne Radioactivity Dose Estimates 12.4-10 for Exposures within the Plant 12.4.3 Exposures at Locations Outside Plant 12.4-11 Structures 12.4.3.1 Direct Radiation Dose Estimates Outside 12.4-12 Plant Enclosures 12.4.3.2 Airborne Radioactivity Dose Estimates 12.4-13 Outside Plant Enclosures 12.4.4 References 12.4-13 12.5 RADIATION PROTECTION PROGRAM 12.5-1 12.5.1 Program Description 12.5-1 12.5.1.1 Authority and Responsibility 12.5-1 12.5.1.2 Experience and Qualifications 12.5-2 12.5.2 Facilities, Equipment, and 12.5-2 Instrumentation 12-iii HCGS-UFSAR Revision 16 May 15, 2008 TABLE OF CONTENTS (Cont)
Section Title Page 12.5.2.1 Radiation Protection and Radiochemistry 12.5-3 Facilities
12.5.2.2 Instruments and Equipment 12.5-7 12.5.3 Procedures 12.5-14
12.5.3.1 Radiological Surveys 12.5-14 12.5.3.2 Radiation Exposure Work Permits 12.5-18 12.5.3.3 Handling and Storage of Radioactive 12.5-19 Material
12.5.3.4 Whole Body Counting 12.5-19 12.5.3.5 Control of Access and Stay Time in 12.5-20 Radiological Areas
12.5.3.6 Radiation Protection Training Programs 12.5-22 12.5.3.7 Radiation Protection Records 12.5-24 12.5.3.8 ALARA Program 12.5-24
12-iv HCGS-UFSAR Revision 16 May 15, 2008 LIST OF TABLES
Table Title
12.1-1 ALARA Plant Model Review - Plant Systems, Areas, and Personnel Activities
12.1-2 ALARA Plant Model Review - Review Documents
12.2-1 Basic Reactor Data
12.2-2 Reactor Vessel and Shield Radial Geometry Used in Calculations to Determine Radial Flux Distribution at Reactor Core Midplane
12.2-3 Material Composition to Determine Radial Flux Distributions at Reactor Core Midplane
12.2-4 Calculated Gamma Ray and Neutron Fluxes at Outside Surface of RPV at Core Midplane
12.2-5 Calculated Gamma Ray and Neutron Fluxes at Outside Surface of Biological Shield at Core Midplane
12.2-6 Reactor Core Fission Product Gamma Sources
12.2-7 Noble Gas Shielding Design Source Terms
12.2-8 Halogen Shielding Design Source Terms
12.2-9 Other Fission Product Shielding Design Source Terms
12.2-10 Coolant Activation Product Shielding Design Source Terms
12-v HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-11 Noncoolant Activation Product Shielding Design Source Terms
12.2-12 Drywell Equipment Drain Sump Shielding Design Source Terms
12.2-13 Drywell Floor Drain Sump Shielding Design Source Terms
12.2-14 RWCU System Recirculation Pumps Shielding Design Source Terms
12.2-15 RWCU System Regenerative Heat Exchanger - First Stage Shielding Design Source Terms
12.2-16 RWCU System Regenerative Heat Exchanger - Second Stage Shielding Design Source Terms
12.2-17 RWCU System Regenerative Heat Exchanger - Third Stage Shielding Design Source Terms
12.2-18 RWCU System Nonregenerative Heat Exchanger - First Stage Tube Side Shielding Design Source Terms
12.2-19 RWCU System Nonregenerative Heat Exchanger - Second Stage Tube Side Shielding Design Source Terms
12.2-20 RWCU System Filter Demineralizer Shielding Design Source Terms
12.2-21 RWCU System Filter Demineralizer Holding Pumps Shielding Design Source Terms
12-vi HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-22 RWCU Backwash Receiver Tank Shielding Design Source Terms
12.2-23 RWCU Backwash Transfer Pump Shielding Design Source Terms
12.2-24 Residual Heat Removal System Shielding Design Source Terms
12.2-25 RCIC System Shielding Design Source Terms
12.2-26 HPCI System Shielding Design Source Terms
12.2-27 Core Spray System Shielding Design Source Terms
12.2-28 Spent Fuel Assembly Shielding Design Source Terms
12.2-29 Fuel Pool Cooling and Cleanup System Shielding
12.2-30 Fuel Pool Filter Demineralizer Shielding Design Source Terms
12.2-31 Fuel Pool Heat Exchangers Shielding Design Source Terms
12.2-32 Traversing In-core Gamma Probe Shielding Design Source Terms
12.2-33 Special Material Sources
12.2-34 Control Rod Drive Mechanism Shielding Design Source Terms
12-vii HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-35 Reactor Building Vent System Filter Shielding Design Source Terms
12.2-36 Primary Containment Prepurge Filter Shielding Design Source Terms
12.2-37 RPV Steam Separator and Dryer Shielding Design Source Terms
12.2-38 Reactor Building Equipment Drain Sump Shielding Design Source Terms
12.2-39 Reactor Building Floor Drain Sump Shielding Design Source Terms
12.2-40 RPV Head Washdown Area Shielding Design Source Terms
12.2-41 Spent Fuel Cask Shielding Design Source Terms
12.2-42 Reactor Building Shielding Design Radiation Source Description
12.2-43 Radwaste Area Equipment Drain Sump Shielding Design Source Terms
12.2-44 Radwaste Area High Conductivity Drain Sump Shielding Design Source Terms 12.2-45 Radwaste Area Floor Drain Sump Shielding Design Source Terms
12-viii HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-46 Floor Drain Collector Tank Shielding Design Source Terms
12.2-47 Floor Drain Collector and Filter Holding Pump Shielding Design Source Terms
12.2-48 Floor Drain Filter Shielding Design Source Terms
12.2-49 Floor Drain Demineralizer Shielding Design Source Terms
12.2-50 Floor Drain Sampling Tank Shielding Design Source Terms
12.2-51 Floor Drain Sampling Pump Shielding Design Source Terms
12.2-52 Detergent Drain Tank Shielding Design Source Terms
12.2-53 Detergent Drain Pump Shielding Design Source Terms
12.2-54 Detergent Drain Filter Shielding Design Source Terms
12.2-55 RWCU Phase Separator Shielding Design Source Terms
12.2-56 RWCU Phase Separator Decant Pump Shielding Design Source Terms
12.2-57 RWCU Sludge Discharge Pump Shielding Design Source Terms
12.2-58 Waste Collector Tank Shielding Design Source Terms
12-ix HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-59 Waste Surge Tank Shielding Design Source Terms
12.2-60 Waste Collector and Surge Pump Shielding Design Source Terms
12.2-61 Waste Filter Shielding Design Source Terms
12.2-62 Waste Demineralizer Shielding Design Source Terms
12.2-63 Waste Sample Tank Shielding Design Source Terms
12.2-64 Waste Sample Pump Shielding Design Source Terms
12.2-65 Chemical Waste Tank Shielding Design Source Terms
12.2-66 Chemical Waste Pump Shielding Design Source Terms
12.2-67 Decontamination Solution Evaporator Shielding Design Source Terms
12.2-68 Decontamination Concentrate Waste Transfer/Recycle Pump Shielding Design Source Terms
12.2-69 Decontamination Solution Concentrate Waste Tank Shielding Design Source Terms
12.2-70 Decontamination Concentrate Waste Pump Shielding Design Source Terms 12.2-71 Waste Neutralizer Tank Shielding Design Source Terms
12-x HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-72 Concentrate Feed/Waste Neutralizer Pump Shielding Design Source Terms 12.2-73 Waste Evaporator Shielding Design Source Terms
12.2.74 Waste Evaporator Condenser Shielding Design Source Terms
12.2-75 Waste Evaporator Distillate Tank Shielding Design Source Terms
12.2-76 Waste Evaporator Distillate Transfer Pump Shielding Design Source Terms 12.2-77 Waste Evaporator Concentrate Transfer/Recycle Pump Shielding Design Source Terms
12.2-78 Concentrate Waste Tank Shielding Design Source Terms
12.2-79 Concentrate Waste Pump Shielding Design Source Terms
12.2-80 Radwaste Cation Regeneration Tank Shielding Design Source Terms
12.2-81 Radwaste Anion Regeneration Tank Shielding Design Source Terms
12.2-82 Spent Resin Tank Shielding Design Source Terms
12.2-83 Spent Resin Pump Shielding Design Source Terms
12-xi HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-84 Radwaste Tank Vent Filter Shielding Design Source Terms
12.2-85 Waste Sludge Phase Separator Shielding Design Source Terms
12.2-86 Waste Sludge Discharge Mixing Pump Shielding Design Source Terms
12.2-87 Piping Between the Aftercondenser and Recombiner System Shielding Design Source Terms
12.2-88 Off-gas Recombiner Shielding Design Source Terms
12.2-89 Off-gas Recombiner Cooler Condenser Shielding Design Source Terms
12.2-90 Off-gas Holdup Decay Pipe Shielding Design Source Terms
12.2-91 Off-gas Charcoal Bed Cooler Condenser Shielding Design Source Terms
12.2-92 Off-gas Guard Bed Shielding Design Source Terms
12.2-93 Off-gas First Charcoal Bed Shielding Design Source Terms
12.2-94 Off-gas Second Charcoal Bed Shielding Design Source Terms
12.2-95 Off-gas Third Charcoal Bed Shielding Design Source Terms
12-xii HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-96 Off-gas Fourth Charcoal Bed Shielding Design Source Terms
12.2-97 Off-gas HEPA Filter Shielding Design Source Terms
12.2-98 Centrifuge Feed Tank Shielding Design Source Terms
12.2-99 Centrifuge Feed Tank Recirculation Pump Shielding Design Source Terms 12.2-100 SRW Centrifuge Shielding Design Source Terms
12.2-101 SRW Extruder Evaporator Shielding Design Source Terms
12.2-102 Volume Reduction Recirculation Pump Shielding Design Source Terms
12.2-103 Crystallizer Bottom Tank Shielding Design Source Terms
12.2-104 Volume Reduction System (VRS) Entrainment Separator Shielding Design Source Terms
12.2-105 VRS Condenser Cooler Shielding Design Source Terms
12.2-106 Radwaste Building Exhaust System Filter Shielding Design Source Terms 12.2-107 Auxiliary Building Shielding Design Radiation Source Description
12-xiii HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-108 Main Steam and Feedwater Heater System Shielding Design Source Terms 12.2-109 Feedwater Heater No. 6 (A, B, or C) Drain Shielding Design Source Terms 12.2-110 Feedwater Heater No. 5 (A, B, or C) Drain Shielding Design Source Terms 12.2-111 Feedwater Heater No. 4 (A, B, or C) Drain Shielding Design Source Terms 12.2-112 Feedwater Heater No. 3 (A, B, or C) Drain Shielding Design Source Terms 12.2-113 Feedwater Heater No. 2 (A, B, or C) Drain Shielding Design Source Terms 12.2-114 Feedwater Heater No. 1 (A, B, or C) Drain Shielding Design Source Terms 12.2-115 Drain Cooler (A, B, or C) Shielding Design Source Terms
12.2-116 Main Condenser Hotwell Shielding Design Source Terms
12.2-117 Primary Condensate Pump Shielding Design Source Terms
12.2-118 Feedwater Shielding Design Source Terms
12.2-119 Condensate Demineralizer Shielding Design Source Terms
12-xiv HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-120 Resin Separation and Cation Regeneration Vessel Shielding Design Source Terms
12.2-121 Anion Regeneration Vessel Shielding Design Source Terms
12.2-122 Ultrasonic Resin Cleaner Shielding Design Source Terms
12.2-123 Turbine Sealing Steam Shielding Design Source Terms
12.2-124 Piping Between the Main Condenser and SJAE System Shielding Design Source Terms
12.2-125 SJAE Intercondenser Shielding Design Source Terms
12.2-126 SJAE Aftercondenser Shielding Design Source Terms
12.2-127 SJAE Intercondenser Drain Shielding Design Source Terms
12.2-128 SJAE Aftercondenser Drain Shielding Design Source Terms
12.2-129 Mechanical Vacuum Pump Shielding Design Source Terms
12.2-130 Turbine Building Equipment Drain Sump Shielding Design Source Terms
12.2-131 Turbine Building High Conductivity Sump Shielding Design Source Terms
12-xv HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title
12.2-132 Turbine Building Floor Drain Sump Shielding Design Source Terms
12.2-133 Turbine Building Shielding Design Radiation Source Description
12.2-134 Post-Accident Source Terms for Drywell and GasContaining Systems
12.2-135 Post-Accident Source Terms for Depressurized Liquid Containing Systems 12.2-136 Post-Accident Source Terms for Pressurized Liquid Containing Systems 12.2-137 Condensate Storage Tank Shielding Design Source Terms
12.2-138 Estimated Inplant Airborne Radioactive Releases (Curies/Year)
12.2-139 Estimated Inplant Distribution of Airborne Radioactivity Releases
12.2-140 Airborne Source Descriptions
12.2-141 Turbine Building Concentrations of MPC Fractions
12.2-142 Reactor Building Concentrations of MPC Fractions
12.2-143 Drywell Concentrations and MPC Fractions
12.2-144 Radwaste Building Concentrations and MPC Fractions
12.2-145 Low Level Radwaste Storage Facility Shielding Design Source Terms
12-xvi HCGS-UFSAR Revision 7 December 29, 1995 LIST OF TABLES (Cont)
Table Title
12.3-1 Plan Area Radiation Zones for Normal Full Power Operation
12.3-2 Post-Accident Shielding Source Terms
12.3-3 List of Post-Accident Vital and Useful Areas
12.3-3a Post-Accident Dose Rates to Vital and other Areas
12.3-4 List of Computer Codes Used in Shielding Design Calculations
12.3-5 Reactor Building Ventilation System Design Features
12.3-6 Turbine Building Ventilation System Design Features
12.3-7 Auxiliary Building Ventilation System Design Features
12.3-8 Control Room Ventilation System Design Features
12.3-9 Area Radiation Monitor Detector Locations
12.3-10 Atmospheric Dispersion factors for Evaluating Post-Accident Access
12.4-1 Summary of Historical Data Used in Compiling Exposures Received at Operating Boiling Water Reactors
12.4-2 Occupational Exposures by Job Function for Operating Boiling Water Reactors
12-xvii HCGS-UFSAR Revision 12 May 3, 2002 LIST OF TABLES (Cont)
Table Title
12.4-3 Summary of Routine Operations Exposure Estimate
12.4-4 Estimate of Expected Routine Maintenance Requirements
12.4-5 Summary of Inplant Direct Radiation Exposure Estimates
12.4-6 Exposure Estimates for Turbine Building Areas
12.4-7 Exposure Estimates for Reactor Building Areas
12.4-8 Exposure Estimates for the Radwaste Areas
12.4-9 Direct Exposure Estimates Due to Routine Surveillance
12.4-10 Estimated Turbine Building Inhalation Exposures Due to Airborne Radioactivity
12.4-11 Estimated Reactor Building Inhalation Exposures Due to Airborne Radioactivity
12.4-12 Estimated Radwaste Areas Inhalation Exposures Due to Airborne Radioactivity
- 12.4-13 Airborne Exposure Estimates Due to Routine Surveillance
12.4-14 Estimated Exposure for Operations in Control Rooms
12.4-15 Skyshine Source Terms
12-xviii HCGS-UFSAR Revision 0 April 11, 1988 LIST OF TABLES (Cont)
Table Title 12.4-16 Direct Radiation Dose Rates at Site Boundary Locations
12.4-17 Direct Radiation Dose Rates at Plant Structures
12.5-1 Deleted 12.5-2 Deleted
12-xix HCGS-UFSAR Revision 22 May 9, 2017
LIST OF FIGURES
Figure Title
12.3-1 Deleted: Refer to Plant Drawing N-1031
12.3-2 Deleted: Refer to Plant Drawing N-1032
12.3-3 Deleted: Refer to Plant Drawing N-1037
12.3-4 Deleted: Refer to Plant Drawing N-1033
12.3-5 Deleted: Refer to Plant Drawing N-1034
12.3-6 Deleted: Refer to Plant Drawing N-1035
12.3-7 Deleted: Refer to Plant Drawing N-1036
12.3-8 Deleted: Refer to Plant Drawing N-1038
12.3-9 Deleted: Refer to Plant Drawing N-1041
12.3-10 Deleted: Refer to Plant Drawing N-1042
12-xx HCGS-UFSAR Revision 20 May 9, 2014 LIST OF FIGURES (Cont)
Figure Title
12.3-11 Deleted: Refer to Plant Drawing N-1043
12.3-12 Deleted: Refer to Plant Drawing N-1044
12.3-13 Deleted: Refer to Plant Drawing N-1045
12.3-14 Deleted: Refer to Plant Drawing N-1046
12.3-15 Deleted: Refer to Plant Drawing N-1047
12.3-16 Deleted: Refer to Plant Drawing N-1011
12.3-17 Deleted: Refer to Plant Drawing N-1012
12.3-18 Deleted: Refer to Plant Drawing N-1013
12.3-19 Deleted: Refer to Plant Drawing N-1014
12.3-20 Deleted: Refer to Plant Drawing N-1015
12.3-21 Deleted: Refer to Plant Drawing N-1016
12-xxi HCGS-UFSAR Revision 20 May 9, 2014 LIST OF FIGURES (Cont)
Figure Title
12.3-22 Control and Diesel. Generator Area Shielding and Radiation Zoning Dwg. Plan el. 155'-3", 163'-6" 12.3-23 Control and Diesel. Generator Area Shielding and Radiation Zoning Dwg. Plan el. 77'-0" 12.3-24 Control and Diesel. Generator Area Shielding and Radiation Zoning Dwg. Plan el. 102'-0" 12.3-25 Control and Diesel. Generator Shielding and Radiation Zoning Dwg.
Plan el. 124'-0", 130'-0",
12.3-26 Control and Diesel. Generator Area Shielding and Radiation Zoning Dwg. Plan el. 137'-0", 146'-0", 150'-0" 12.3-27 Control and Diesel. Generator Area Shielding and Radiation Zoning Dwg. Plan el. 178'-0" 12.3-28 Control and Diesel. Generator Area Shielding and Radiation Zoning Dwg. Plan el. 54'-0" 12.3-29 Site Plan Shielding and Radiation Zoning Drawing 12.3-30 Deleted
12.3-31 Deleted
12.3-32 Deleted
12-xxii HCGS-UFSAR Revision 12 May 3, 2002 LIST OF FIGURES (Cont)
Figure Title
12.3-33 Deleted
12.3-34 Deleted
12.3-35 Deleted
12.3-36 Deleted
12.3-37 Deleted
12.3-38 Deleted
12.3-39 Deleted
12.3-40 Deleted
12.3-41 Deleted
12.3-42 Deleted
12-xxiii HCGS-UFSAR Revision 12 May 3, 2002 LIST OF FIGURES (Cont)
Figure Title
12.3-43 Deleted
12.3-44 Deleted
12.3-45 Deleted
12.3-46 Deleted
12.3-47 Deleted
12.3-48 Deleted
12.3-49 Deleted
12.3-50 Deleted
12.3-51 Deleted
12.3-52 Deleted
12-xxiv HCGS-UFSAR Revision 12 May 3, 2002 LIST OF FIGURES (Cont)
Figure Title
12.3-53 Deleted
12.3-54 Deleted
12.3-55 Deleted
12.3-56 Deleted
12.3-57 Deleted
12.3-57a Deleted
12.3-58 Typical Shielding Arrangement for Charcoal and Particulate Filter
12.3-59 Functional Block Diagram Area Radiation Monitoring System
12.3-60 Equipment Location Turbine Building Plan el. 137'-0"
12.3-61 Equipment Location Turbine Building Plan el. 102'-0"
12.3-62 Equipment Location Turbine Building Plan el. 77'-0"
12.3-63 Equipment Location Turbine Building Plan el. 54'-0"
12-xxv HCGS-UFSAR Revision 12 May 3, 2002 LIST OF FIGURES (Cont) Figure Title 12.3-64 Deleted: Refer to Plant Drawing P-0047-1 sheets 1 and 2 12.3-65 Equipment Location Reactor Building Plan el. 145'-0" 12.3-66 Deleted: Refer to Plant Drawings P-0044-1 sheet 1 and P-0046-1 sheet 1 12.3-67 Equipment Location Reactor Building Plan el. 102'-0" 12.3-68 Equipment Location Turbine Building Plan el. 77'-0" 12.3-69 Equipment Location Service and Radwaste Area Plan el. 102'-0" 12.3-70 Equipment Location Service and Radwaste Area Plan el. 124'-0" 12.3-71 Equipment Location Service and Radwaste Area Plan 54'-0" 12.4-1 Isodoses and Receptor Point Exposures 12.5-1 Figure Deleted 12.5-2 Deleted: Refer to Plant Drawing P-0035-0 12.5-3 Access Control - Radiation Protection and Chemistry Facilities 12.5-4 This Figure has been removed. See the latest revision of drawing A-B102-0 Sheet 001 12.5-5 This Figure has been removed. See the latest revision of drawing A-B103-0 Sheet 001 12-xxvi HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY
ACHIEVABLE
12.1.1 Policy Considerations 12.1.1.1 Management Policy
The policy of PSEG Nuclear LLC is to maintain occupational radiation exposures "as low as reasonably achievable" (ALARA) at Hope Creek Generating Station (HCGS). This includes maintaining the annual integrated dose to station personnel and to individuals working at the station ALARA. PSEG management is firmly committed to performing all reasonable actions to ensure that radiation
exposures are maintained ALARA.
Sections 12.1.2, 12.3.1, and 12.3.2 discuss the ALARA considerations that have been incorporated into the design of HCGS.
HCGS is operated and maintained in such a manner as to ensure occupational radiation exposures are ALARA. The operational ALARA program is described in
Section 12.5. Training programs ensure that personnel understand both how and why occupational radiation exposures are maintained ALARA. A corporate AL ARA program ensures implementation of the station ALARA policy by various program
reviews.
12.1.1.2 Management Responsibilities
Figures 13.1-1 through 13.1-4 show the PSEG Nuclear management organizational structure for HCGS.
The President and Chief Nuclear Officer (P/CNO) has the corporate responsibility for the ALARA program. The responsibility for coordination and administration of the ALARA program is delegated to the Radiation Protection Manager who is responsible for ensuring that radiation protection policies and
12.1-1 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information)
12.2 RADIATION SOURCES
This section identifies and discusses the sources of radiation that form the basis for the radiation shielding design required for in-plant radiation protection, for the plant ventilation design to control airborne radioactive contamination, and for dose assessment. Sources due to normal full power
operations and accident conditions are addressed.
The nuclear fission process in the reactor core produces a large number of radiation sources. Some of these sources are stationary and remain within the reactor vessel and its internal and external structures. Other radiation sources are released and transported in the form of radioactive fission and activation products. The transport and subsequent distribution of these sources is by way of the Reactor Coolant System (RCS), the Main Steam Supply System, and the numerous auxiliary plant support systems. Following an
accident, additional and much stronger sources could be released to the RCS, the engineered safeguard systems, and throughout large volumes of the primary
12.2.1 Contained Sources
The shielding design source terms for normal full power plant operations are based on a noble gas fission product release rate of 0.50 Ci/s (after 30 minutes of decay) and the corresponding fission, activation, and corrosion product concentrations in the primary coolant. The guidance provided in ANSI N237 was not followed for Hope Creek Generating Station (HCGS). Acceptance Criteria II.6 of SRP Section 12.2 addresses the use of ANSI Standard N237-1976 "Source Term Specification," and its use in establishment of a typical long term concentration of principal radionuclides in fluid streams. GE developed and upgraded source terms based on operating plant experience. These specifications are more applicable to boiling water reactors (BWRs). The specific alternate methods and design basis used for calculating source terms
and their magnitudes are described in Section 11.1.
12.2-1 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
Throughout most of the primary coolant and main steam systems, activation products, principally N 16, are the primary radiation sources for the shielding design.
Other activation products, such as Co 60 , and certain fission products become important as built-up or accumulation sources in filters, demineralizers, or in plateout processes at pipe and equipment surfaces. The majority of these sources remain contained within the confines of the closed plant piping systems
and the various tank and container facilities.
Basic reactor data and core region descriptions used for this section are
listed in Section 12.2.1.1.1.
The shielding design radiation source terms are presented by building, location, and system. Locations of the equipment discussed in this section are shown on the shielding and radiation zoning drawings, Figures 12.3-22 through 12.3-29 and Plant Drawings N-1031 through N-1038, N-1041 through N-1047 and N-1011 through N-1016. Detailed data on source descriptions for shielded areas
are presented in Tables 12.2-42, 12.2-107, and 12.2-133.
Shielding source terms presented in this section, and associated tables, are based on conservative assumptions for the system and equipment operations and characteristics to provide conservative radioactivity concentrations for the shielding design. For all systems transporting radioactive materials, conservative allowance is made for transit decay while at the same time
providing for daughter product formation.
12.2.1.1 Primary Containment (Drywell)
12.2.1.1.1 Reactor Core
This section provides information needed to form a reactor core source model.
It also provides multi-group neutron and gamma ray fluxes at important locations at and near the surface of the reactor vessel. This data is required
for the calculations
12-2-2 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information)
determining the primary radiation shielding requirements for the drywell structures, and to establish the radiological environmental conditions inside
the primary containment.
Table 12.2-1 presents physical data needed in forming a reactor core source model. The data includes the core thermal power, the power peaking factors (peak to average at core center and core boundary to average), volume fractions
for the core, and the water density conditions in the non-core regions.
Table 12.2-2 presents the reactor vessel radial geometry data used in calculations to determine the radial radiation flux distributions at the core
mid-plane.
Table 12.2-3 gives the material compositions used in calculating the radial radiation flux distributions at the core mid-plane. Tables 12.2-4 and 12.2-5 present calculated gamma ray and neutron fluxes at core mid-plane at the outside surface of the reactor pressure vessel (RPV), and at the outside surface of the biological shield, respectively. The gamma fluxes include those resulting from capture or inelastic scattering of neutrons within the RPV and core shroud, and the gamma radiation resulting from prompt fission and fission product decay. The calculated gamma ray fluxes in this section do not include
any provisions for scattering from points outside of the vessel. Also, there
are no provisions for gamma ray fluxes from deposits of radioactive isotopes
within the vessel or from the neutron activation of reactor vessel materials.
The largest radiation sources after reactor shutdown are the decaying fission products in the fuel. Table 12.2-6 lists the reactor core fission product gamma sources at 3 days after shutdown. Secondary sources are the structural material activation of the RPV, its internals, and also the activated corrosion
products accumulated or deposited in the internals of the RPV.
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12.2.1.1.2 Reactor Coolant System
Sources of radiation in the Reactor Coolant System (RCS) are both fission products released from a corresponding percentage of defective fuel, based on a noble gas fission product release rate of 0.50 Ci/s (after 30 minutes of decay), and activation and corrosion products that are circulated in the reactor coolant. These sources are listed in Tables 12.2-7 through 12.2-11.
The N 16 concentration in the reactor coolant is assumed to be 40 Ci/g of coolant at the reactor recirculation outlet nozzle.
12.2.1.1.3 Main Steam System
Radiation sources in the Main Steam System piping include activation gases, principally N 16, and the activated corrosion products and fission products carried over to the steam system.
The N 16 concentration in the main steam is assumed to be 50 Ci/g of steam leaving the reactor vessel at the main steam outlet nozzle. The fission product activity for the purpose of personnel shielding design corresponds to an off-gas release rate of 500,000 Ci/s at 30 minutes decay from the reactor steam nozzle. Carryover fractions for activity into the steam system are 100 percent for the gaseous fission and activation products. Carryover of radioiodines from the reactor coolant water to the steam is taken as 2 percent by weight of the reactor water. Carry-over of other radioisotopes is taken as 0.1 percent by weight of the reactor water. Main steam radiation sources are
shown in Tables 12.2-7 through 12.2-11.
12.2.1.1.4 Drywell Sumps
The concentrations of radioisotopes used for shielding design for the drywell
equipment and floor drain sumps are listed in Tables 12.2-12 and 12.2-13.
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12.2.1.2 Reactor Building
12.2.1.2.1 Reactor Water Cleanup System
Radiation sources in the Reactor Water Cleanup (RWCU) System consist of those radioisotopes carried in the reactor coolant water. The activity inventory is based on reactor coolant water radioactivities corrected for transit times to
the individual components of the RWCU system.
The radioisotopes for the RWCU recirculation pumps, regenerative and nonregenerative heat exchangers, and associated piping are based on the inlet reactor coolant concentrations given in Tables 12.2-14 through 12.2-19, allowing for decay due to transit time.
Radioisotope concentrations in the filter demineralizers, holding and transfer pumps, and the backwash receiving tank are based on the accumulation of fission and activation products. Tables 12.2-20 through 12.2-23 provide the shielding
design source terms for these components.
12.2.1.2.2 Residual Heat Removal System
The radioactive sources in the Residual Heat Removal (RHR) System are evaluated for the system operating in the reactor shutdown mode. In this mode, the system recirculates reactor coolant to remove reactor core decay heat. The system is described in Section 5.4.7. The system is operated from approximately four hours after shutdown until the end of the refueling period.
The radiation source in the RHR system is based on the maximum estimated
activity due to the iodine spike effect at four hours of decay following shutdown. The source terms listed in Table 12.2-24 are used for the shielding
calculations for this system.
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12.2.1.2.3 Reactor Core Isolation Cooling System
Components of the Reactor Core Isolation Cooling (RCIC) System that contain radiation sources during normal power operation are the RCIC turbine steam inlet and exhaust piping. The steam radioactivity, as discussed in Section 12.2.1.1.3, decayed for the appropriate transit time, is used for the
shielding calculations for this system and is listed in Table 12.2-25.
12.2.1.2.4 High-Pressure Coolant Injection System
The radiation sources for the High Pressure Coolant Injection (HPCI) System are the HPCI turbine steam inlet and exhaust piping. The steam radioactivity, as discussed in Section 12.2.1.1.3, decayed for the appropriate transit time, is used for the shielding calculations for this system and is listed in
Table 12.2-26.
12.2.1.2.5 Core Spray System
During testing, the Core Spray System components use condensate from the condensate storage tank or draw water from the suppression pool. Source terms based on suppression pool activities listed in Table 12.2-27 are used for
shielding the core spray equipment.
12.2.1.2.6 Spent Fuel Storage and Transfer Sources
The predominant radiation sources in the spent fuel storage and transfer areas are the spent fuel assemblies and activated control rods. For shielding design purposes, the spent fuel pool is assumed to contain 3667 fuel assemblies. These spent fuel assemblies are assumed to have 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay before being stored. Shielding design source terms for spent fuel assemblies are shown in Table 12.2-28. The shielding design source terms have been verified to bound the source term for the expected storage of spent fuel and irradiated
components. The shielding design source terms will remain bounding up to the
inclusion of 6241 storage locations in the spent
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fuel storage area. The storage locations can hold fuel assemblies in an as-built condition or in a consolidated configuration. The spent fuel assemblies are assumed to have 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay before being stored in their as-built condition and 14 days of decay before being stored in a 2 to 1 consolidated
configuration.
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THIS PAGE INTENTIONALLY BLANK
12.2-6b HCGS-UFSAR Revision 5 May 11, 1993 (Historical Information)
12.2.1.2.7 Fuel Pool Cooling and Cleanup System
Sources in the Fuel Pool Cooling and Cleanup (FPCC) System are primarily a
result of a transfer of radioactive isotopes from the reactor coolant into the spent fuel pool during refueling operations. The reactor coolant activities for fission, corrosion, and activation products (Tables 12.2-7 through 12.2-11) are decayed for the amount of time required to remove the reactor vessel head
following shutdown, are reduced by operation of the RWCU system filter demineralizers following shutdown, and are diluted by the total volumes of the water in the reactor vessel, refueling pool, and spent fuel pool. See Table 12.2-29. This activity then undergoes subsequent accumulation and decay on the FPCC filter demineralizers. The FPCC filter demineralizer resins are backwashed periodically into the waste sludge phase separator. Sources for FPCC heat exchangers and system piping are based on crud plateout on these components. Table 12.2-30 provides the FPCC filter demineralizer shielding design source terms, and Table 12.2-31 provides the shielding design source
terms for the FPCC heat exchangers and associated piping.
12.2.1.2.8 Radiation Sources in the Traversing In-core Gamma Probe System
The radiation source data for the traversing in-core gamma probe (TIP) system
is provided in Table 12.2-32.
12.2.1.2.9 Reactor Startup Sources
The reactor startup sources are shipped to the site in a specially designed shielding cask. The sources are transferred underwater from the cask and loaded into beryllium containers. This is then loaded into the reactor while remaining underwater. The sources remain within the reactor for their lifetime. Thus, there are no unique shielding requirements after reactor operation. Table 12.2-33 gives information on source strength and other
important data.
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12.2.1.2.10 Control Rod Drives (CRDs)
Shielding design source terms for the control rod drives (CRDs) after removal
from the RPV are shown in Table 12.2-34.
12.2.1.2.11 Reactor Building Ventilation Systems
The Reactor Building Ventilation System (RBVS) exhaust air filters, as part of the Reactor Building Heating, Ventilation, and Air Conditioning (HVAC) Systems, contain sources of radioactivity. Table 12.2-35 shows the RBVS exhaust air filter shielding design source terms. The shielding design source terms for
the primary containment prepurge filters are given in Table 12.2-36.
12.2.1.2.12 Reactor Vessel Steam Separator and Steam Dryer
Shielding design source terms for the steam separator and steam dryer are given in Table 12.2-37. These shielding design source terms are based on estimated
crud plateout activity from operating plants.
12.2.1.2.13 Reactor Building Sumps
The concentrations of radioisotopes used for the shielding design for the Reactor Building Equipment and floor drain sumps are listed in Tables 12.2-38
and 12.2-39.
12.2.1.2.14 RPV Head and Spent Fuel Cask
The shielding design source terms for the RPV head washdown area are shown in Table 12.2-40. Shielding design source terms for the spent fuel shipping cask
are shown in Table 12.2-41.
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12.2.1.2.15 Reactor Building Source Descriptions
A listing and detailed description of the shielding design radiation sources
for the Reactor Building are shown in Table 12.2-42.
12.2.1.3 Service and Radwaste Areas of the Auxiliary Building
The numerous plant radioactive wastes are collected and processed, safely inside the service and radwaste areas of the auxiliary building. Radioactive wastes are in liquid, gaseous, and solid phases. Each type of radwaste, and its associated radiation sources, is contained and handled in a specific manner to ensure maximum radiation protection and minimum plant contamination.
Detailed system and equipment descriptions are provided in Sections 11.2, 11.3, and 11.4.
12.2.1.3.1 Liquid Radwaste System
The liquid radwastes are collected and contained inside the service and radwaste areas from the following systems: fuel pool cleanup, equipment drain, floor drain, chemical waste, condensate, decontamination, and reactor water cleanup. Further processing will be accomplished in the solidification and volume reduction system discussed in Section 12.2.1.3.3. The liquid radwaste
system shielding design sources are radioisotopes, including fission and corrosion products, present in the reactor coolant. The components of the Liquid Radwaste System contain varying amounts of radioactivity, depending on
the origin of the handled radwaste and the system and equipment design.
The concentrations of radioisotopes used for the shielding design for pipes, pumps, tanks, filters, demineralizers and evaporators, and equipment and floor drain sumps are listed in Tables 12.2-43 through 12.2-86. Shielding for each component of the Liquid Waste Management System is based on reactor coolant
radioactivity concentrations given in Tables 12.2-7 through 12.2-11.
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12.2.1.3.2 Off-gas Treatment System
Radioactive off-gases are contained and processed in the Off-gas Treatment System. The major radiation sources of this system, located in the service and radwaste areas of the Auxiliary Building, are found in the off-gas recombiner
cells and the Off-gas Charcoal Treatment System.
The shielding design radiation source terms for the recombiner and charcoal treatment system components are based on the expected transit times for N 16 , noble gases, and the formation and accumulation of noble gas daughter products. Eighty percent of the N 16 and 100 percent noble gases are assumed to be removed from the main condensers by the steam jet air ejectors. These gases pass from the steam jet air ejector to the recombiner equipment and to the charcoal treatment system. The charcoal treatment system functions to filter, as well as to delay, the release of the radioactive offgases. Off-gas release to the environment is through the plant north vent. The shielding design source terms for the piping, recombiner components, and charcoal treatment
system equipment are presented in Tables 12.2-87 through 12.2-97.
12.2.1.3.3 Deleted
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(Historical Information)
12.2.1.3.4 Service and Radwaste Area Ventilation Systems
Components of the ventilation system inside the service and radwaste areas that contain sources of radioactivity are the radwaste exhaust and the Reactor Building vent system filters. Tables 12.2-106 and 12.2-35 show the exhaust air
filter shielding design radiation source terms.
12.2.1.4 Turbine Building
12.2.1.4.1 Main Steam and Power Conversion Systems
Radiation sources for piping and equipment that contain main steam are based on the radioisotopes carried into the main steam from the reactor coolant. The
sources include fission product gases and halogens, particulate fission and corrosion products, and gaseous activation products as discussed in Section 12.2.1.1.3. Steam density variations and steam transit times through equipment and pipes are factored into the shielding source term evaluation to account for volumetric dilution effects, radiological decay, and daughter product generation. Tables 12.2-108 through 12.2-115 show the shielding design
radiation source terms for the following major components that contain main steam as the dominant source: moisture separators, crossaround piping, feedwater heaters, drain
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coolers, steam seal evaporator, reactor feed pump turbines (RFPT), and the
steam packing exhaust condenser.
12.2.1.4.2 Condensate and Feedwater Systems
The radiation sources in the condensate and feedwater systems are based on decayed main steam radioactivity. See Section 12.2.1.1.3. Eighty percent of the N 16 and 100 percent of the noble gases are assumed to be removed from the condensate and feedwater systems by the main condenser air removal system. The radiation sources in the hotwell are shown in Table 12.2-116. They are
negligible as radiation sources in the remainder of the condensate and feedwater systems. The hotwell is designed for a 3-minute holdup of condensate, and therefore N 16 radioactivity at the condenser outlet is negligible. Particulate fission products, activated corrosion products, and the particulate daughter products from the decay of fission product gases in
transit through the turbine and condenser are the inlet radiation sources to the condensate system. Tables 12.2-117 through 12.2-123 provide the shielding
design radiation source terms for the condensate pumps and their associated piping, the condensate filter demineralizers, the resin mix and hold tank, the anion and cation regeneration tanks, the ultrasonic resin cleaner, and the
shielding design radiation source terms for the feedwater system.
12.2.1.4.3 Steam Jet Air Ejector System
The steam jet air injectors (SJAEs) are the only major equipment of the plant Off-gas Treatment System located inside the Turbine Building. Shielding design radiation sources in the Off-gas Treatment System are developed from noble gases and other noncondensable gases removed from the main condenser, and the radioactivity entering with the extraction steam to the SJAEs. The extraction steam radioactivity entering is based on the main steam radioactivity, as described in Section 12.2.1.1.3, and decayed for the expected transit time to the SJAE system. Eighty percent of the N 16 and 100 percent of the noble gases are assumed to be
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Provisions have been made to recycle the water from the CST through the condensate demineralizers.
The third source of radioactivity not stored inside the plant structure is the Independent Spent Fuel Storage Installation (ISFSI). The ISFSI is operated under the general license provisions of 10 CFR 72, Subpart K. It provides additional on-site storage for HCGS spent nuclear fuel removed from the spent fuel pool stored in dry, seal welded stainless steel storage canisters. Each storage canister is located inside a ventilated concrete overpack at the ISFSI that provides structural protection, decay heat removal, and shielding. The storage canisters are considered leak-tight under all design basis normal, off-normal, and accident conditions as described in the spent fuel storage system FSAR. Therefore, there are no effluent releases from the dry storage systems at the ISFSI. Source terms and other design and operating information for the dry spent fuel storage system can be found in the storage system 10 CFR 72 FSAR. A discussion of the off-site dose contribution due to direct radiation from the ISFSI may be found in the 10 CFR 72.212 evaluation report for the
Under normal conditions, no other radiation sources are stored outside plant structures without prior approval of the radiation protection engineer. Spent fuel is stored in the spent fuel pool until it is placed in the spent fuel shipping cask for offsite transport. All radiation sources contained inside the plant structures are shielded to provide a dose rate of less than
0.5 mrem/h for all areas outside plant structures.
(Historical Information)
12.2.1.7 Special Sources
Special radioactive materials used in the radiochemistry laboratory, and sealed sources used for calibration, require special handling equipment and are shielded accordingly. A listing of these special material sources can be found in Table 12.2-33. Unsealed sources and radioactive samples are handled under conventional hoods that exhaust to the plant ventilation systems. Design
features are discussed in Section 12.3.1.
12.2-14 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 12.2.2 Airborne Radioactive Material Sources
12.2.2.1 Sources of Airborne Radioactivity
The sources of airborne radioactivity found in the various areas of the plant are mostly from process leakage of the systems carrying radioactive gases, steam, and liquids. Depending on the type of system and its physical conditions, such as system pressures and temperatures, leakage is in the form
of a gas, steam, liquid, or a mixture of these.
12.2.2.2 Emission of Airborne Radioactive Materials
Radioactive materials become airborne through a number of mechanisms. The most common emission mechanisms are spraying, splashing, flashing, evaporation, and
diffusion.
12.2.2.3 Locations of Sources of Airborne Radioactivity
Practically all sources of airborne radioactivity are found in the Reactor and Turbine Buildings and the service and radwaste areas of the Auxiliary Building.
Within these structures, the radioactivity is released in equipment cubicles, valve and piping galleries, sampling stations, radwaste handling and shipping areas, cleaning and decontamination areas, fuel transfer and pool storage areas, and repair shops. Insignificant amounts of airborne radioactivity could occur in places such as the radiation protection work area or the condensate
storage tank area.
12.2.2.4 Control of Airborne Radioactivity
Ventilation and filtration are the most effective means of controlling airborne
radioactive materials. Ventilation flow paths are designed so that air from areas of low potential airborne radioactivity flows into areas with higher potential for airborne radioactivity. Such a flow pattern ensures that radioactivity released in the source locations mentioned above has a low probability to escape into areas with high personnel occupancy requirements, such as corridors, working aisles, and operating floors. Levels of airborne radioactivity are continuously monitored by the area radiation monitor and, the
airborne radiological monitoring systems, and they are also periodically
checked through surveys of the plant by the radiation protection staff.
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12.2.2.5 Methodology for Estimating the Concentration of Airborne Radioactive Material Within the Plant
To estimate the airborne radioactive material concentrations at locations
within the plant, the following methodology is used:
- 1. Estimate the total airborne releases (in Ci/yr) for each of the plant enclosures.
- 2. Estimate a distribution for these releases among the various equipment areas of each enclosure based on operating data and
engineering judgment.
- 3. Determine the annual exhaust flow from each equipment area.
- 4. Calculate the resultant airborne radionuclide concentration (Ci/cc) in each equipment area based on the release distribution (Ci/yr)
and exhaust flow rate (cc/yr).
The following sections discuss each step in the procedure listed above in more
detail.
12.2.2.5.1 Estimation of Total Airborne Releases Within the Plant During Normal Full Power Operations The estimated quantities of airborne radioactive material produced in the plant enclosures are given in Table 12.2-138. These releases were based upon BWR gaseous and liquid effluents, mentioned in Reference 12.2-1, using a computerized mathematical model (BWR-GALE) for calculating the release of radiological materials in gaseous and liquid effluents. Assumptions applicable
to the development of Table 12.2-138 from BWR-GALE are as follows:
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- 1. The drywell and torus releases are taken to be the Containment Building releases calculated by BWR-GALE.
- 2. The releases from the HCGS Reactor Building correspond to the Auxiliary Building releases calculated by BWR-GALE.
- 3. Turbine Building releases from BWR-GALE are assumed to include any airborne radioactive material released in the individual rooms and
cubicles of the structure.
- 4. The radwaste area releases from BWR-GALE also include offgas system releases.
- 5. Tritium releases from BWR-GALE are divided equally between the reactor and turbine buildings.
- 6. Since the BWR-GALE code for gaseous releases is based on actual operating plant data, releases for both normal operations and
anticipated operational occurrences are included.
12.2.2.5.2 Distribution of Airborne Releases Within the Plant
In the approach taken to determine the anticipated distribution of gaseous effluents, it is assumed that all airborne radioactive material originates only
within the equipment areas of the plant.
It is further assumed that a major percentage of the release is generated within a few specific areas of each enclosure, with the remainder coming from the "remaining equipment areas." Eighty percent of each enclosure's release is distributed, as described below, among the major contributing areas, and 20 percent is assigned to the remaining equipment areas category. Releases are assumed to be generated continuously throughout the year, except for the
drywell, where a 30-day annual release period is used.
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The basis for the selection and relative contributions of the major areas is the Electric Power Research Institute Report, EPRI NP-495, Reference 12.2-2. This report provides data on the important sources of I 131 at operating BWRs and uses measured data to determine the relative release rate from each source.
The relative release rates for all airborne radionuclides, except tritium from the Reactor Building, are assumed to be directly proportional to the I 131 release rates. Since the spent fuel pool and the reactor vessel (when it is open during refueling) are the major sources of airborne tritium in the Reactor Building, tritium releases for that enclosure are assigned entirely to the
refueling area.
Table 12.2-139 lists the major airborne contributors in each enclosure and the percentage of the total enclosure release assigned to each. Table 12.2-140 also provides the specific equipment areas of the plant associated with the major contributors and the applicable exhaust air flow rates. Note that only those equipment areas that have a significant potential for airborne
radioactive material releases are included in the remaining equipment areas
category.
12.2.2.5.3 Estimated Airborne Radioactive Material Concentrations Within the Plant The airborne radionuclide concentrations for each equipment area are calculated for a specific area, by multiplying the appropriate enclosure release, Table 12.2-138, by the applicable release percentage for the area, Table 12.2-139, and dividing that by the area's annual exhaust flow, Table 12.2-140. The resultant concentrations are presented in Tables 12.2-141 through 12.2-144, which also include the fractions of the maximum permissible concentrations in
air as defined in 10CFR20, Appendix B, Table I, Reference 12.2-3.
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12.2.2.5.4 Changes to Source Data Since PSAR
Liquid, solid, and airborne radioactive material sources are not specified in the HCGS PSAR. Section 12.2 has been added to the HCGS FSAR in compliance with
Regulatory Guide 1.70, Revision 3, Reference 12.2-4.
12.2.3 References
12.2-1 Nuclear Regulatory Commission, Office of Standard Development, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents From Boiling Water Reactors," NUREG-0016, Revision 1, January 1979.
12.2-2 Electric Power Research Institute, "Sources of Radioiodine at Boiling Water Reactors," EPRI NP-495, Project 274-1, Final Report, February 1978.
12.2-3 Nuclear Regulatory Commission, "Standards for Protection Against Radiation" "Code of Federal Regulations," Title 10, Part 20, November 1975.
12.2-4 Nuclear Regulatory Commission, "Standard Format and Content of Safety Analysis Report for Nuclear Power Plants," Regulatory
Guide 1.70, Revision 3, November 1978.
12.2-5 Nuclear Regulatory Commission, "Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980.
12.2-6 251 Standard Safety Analysis Report, General Electric, for BWR/6, Revision 22.
12.2-19 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
12.2-7 Technical Information Document, TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites by J.J. DiNunno, R.E. Baker, F.D. Anderson and R.L. Waterfield, March 23, 1962.
12.2-20 HCGS-UFSAR Revision 16 May 15, 2008
(Historical Information) 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 Facility Design Features Specific design features for maintaining personnel exposures as low as is reasonably achievable (ALARA) are discussed in this section. These features are used in addition to and in conjunction with the more generalized design features described in Section 12.1.2. 12.3.1.1 Common Equipment and Component Layout and Designs for ALARA This section describes the design features used for several general classes of equipment and components. Since these classes of equipment are common to many of the plant systems, the features employed in each system to minimize exposures are similar and can be discussed generically by equipment type. 1. Filters - Filters that accumulate a significant amount of radioactivity are supplied with the means either to backflush and recharge the filter remotely or to perform cartridge replacement with semi-remote tools, i.e., long handled tools. For cartridge filters, adequate space is provided to allow removing, loading, and transporting the cartridges to the solid radwaste storage area in a shielded cask or drum. To make the removal of cartridges easier, quick disconnect features are provided to minimize the time needed for personnel to open/close the filter housings. Filters that are expected to accumulate very high amounts of radioactivity are provided with shielded remote reading instrument plug openings in their concrete ceilings. Remote reading instruments allow semi-remote readings of filter surface dose rates for radiation surveys without the opening and entering of filter cubicles and aid plant personnel in planning and preparation of radiation protection features. 12.3-1 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 2. Demineralizers - Demineralizers in radioactive waste and water treatment systems are designed so that spent resin can be remotely transferred to spent resin or regeneration tanks, and fresh resin can be remotely loaded into the demineralizers. The demineralizers and piping can be flushed with condensate or demineralized water to remove any accumulation of resin. The equipment and floor drain filter/demineralizers are typical examples, shown on Plant Drawings M-62-0 and M-63-0, respectively. The cubicles for the condensate demineralizers expected to accumulate very large amounts of radioactivity are provided with a shielded remote reading instrument plug opening in their concrete ceilings. Remote reading instruments allow semiremote readings of demineralizer surface dose rates for radiation surveys without opening and entering demineralizer cubicles. They also aid in planning and preparation of shielding features to provide for cubicle entry, if necessary. 3. Evaporators - Evaporators are provided with chemical addition and condensate or demineralized water flushing connections that also allow the use of chemicals for decontamination operations shown on Plant Drawing M-64-0. Adequate space is provided to facilitate removal of evaporator tube bundles. High activity concentrate carrying and low activity distillate carrying components are separated from each other wherever possible. 4. Pumps - Wherever practicable, pumps in radioactive areas are provided with mechanical seals to reduce seal servicing time. Pumps and associated piping are arranged to provide adequate space and access for maintenance. 12.3-2 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) Small pumps are installed in a manner that allows easy removal if necessary. All pumps in radioactive waste and water treatment systems are provided with flanged connections for ease of removal. Pump casings and pump associated pipe spools have connections for draining the pump and pipe sections prior to maintenance. The use of base plates with drains connected to the floor drain system minimizes the spread of contamination resulting from pump leakage. Where possible, pumps in radioactive areas are placed in individual cubicles to enhance radiation protection and minimize the spread of contamination. 5. Tanks - Whenever practicable, tanks that contain radioactive material are provided with sloped bottoms and bottom outlet connections. Overflow lines are lower than vent lines and are directed to the waste collection system to prevent an overflow from spreading contamination within plant structures. For tanks containing radioactive material, each tank can be rinsed, and associated piping can be flushed to reduce radiation levels if required for entry into the tank cells. Adequate access and headroom are provided for removal, maintenance, or inspection of tank motor agitators and also for cleaning operations involving tank internals. The Solid Radwaste Collection System shown on Plant Drawings M-66-0, M-67-0 and M-68-0 provides examples of these features. All radioactive tanks are placed in individual, shielded cubicles. Active components such as valves are placed outside the cubicles to the extent possible. Those tanks holding radioactive liquids are surrounded by walls or dikes providing containment in case of a tank rupture. The diked-in areas are provided with floor drains for normal low flow leakage. 12.3-3 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) 6. Heat exchangers - Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials with tube-to-tube sheet joints, welded to minimize leakage. Impingement baffles are provided, and tube side and shell side velocities are limited to minimize erosive effects. Space is reserved for tube bundle removal, and provisions are made for drainage and flushing. 7. Turbine condenser - Most of the condenser instrument readout devices are located in a shielded and low radiation condenser instrument compartment. The condenser is provided with corrosion resistant tubes of titanium-alloy steel to minimize corrosion effects and, in turn, reduce the need for tube plugging or other related maintenance and repair activities. Radioactive pipes have been routed in such a way as to minimize radiation exposures to personnel working at or near the condenser. Permanent ladders and platforms are provided to further enhance plant personnel maintenance operations and reduce radiation exposures. 8. Instruments - Instrument transmitting and readout devices are located in low radiation zones and away from radiation sources whenever practicable to reduce personnel exposure during maintenance. Sensing instrument devices, which for functional reasons are located in high radiation zones, are designed for easy removal to a lower radiation zone for maintenance and calibration. Some instruments located in high radiation zones, such as thermocouples, are provided in duplicate to preclude the need for immediate entry in case of instrument failure and to allow maintenance to be performed at a later time when radiation levels may be lower. Instrument and sensing lines are provided with flushing capability and are routed to minimize radioactive crud buildup. Backflushing capability exists for reactor vessel sensing lines. Tanks containing two phase fluids are fitted with probe type instruments. Instrument and sensing line connections are typically located at or above the piping midplane to avoid corrosion product buildup. 12.3-4 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 9. Valves - To minimize personnel exposures from valve operations and maintenance, motor operated, air operated, or other remotely actuated valves are used to the maximum extent practicable. Savings in personnel exposures will result from the large reductions of operational exposures compared to those from the maintenance of remote-operator components. Whenever practicable, valves are located in valve galleries and are shielded separately from the major components that accumulate radioactivity. Long runs of exposed radioactive piping are minimized in valve galleries. In areas where manual valves are used on frequently operated radioactive process lines, radiation shielding is provided to minimize personnel exposure. No valves or other active components are placed in radioactive pipe chases. For equipment located in high radiation zones, remote actuators are provided for frequently operated valves associated with system operation. All other valve operations are either infrequent or are performed when the equipment is not operating. Provisions are made in many radiation areas to drain and flush adjacent radioactive components when maintenance is required on valves. 12.3-5 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) Wherever practicable, valves for clean, non-radioactive systems are separated from radioactive sources and are located in readily accessible areas. Vent, drain, and instrument root isolation valves on radioactive systems are located close to the process piping or equipment with which they are associated. This minimizes the lengths of piping carrying process fluids when these valves are closed. These valves are generally manually operated since they are used infrequently. Where possible, valves required for normal operation and shutdown are not located in filter and demineralizer compartments. For large valves (2-1/2 inches and larger) in lines carrying radioactive fluids, a double set of packing with a lantern ring is usually provided. Full ported valves are used in systems containing radioactive slurries. For the Radioactive Off-gas System, zero stem leakage valves are used to minimize the spread of airborne radioactive contamination. Valve designs with minimum internal crevices are used where trapping crud could become a problem, such as in piping carrying spent resin or evaporator bottoms. 10. Floor drains - Floor drains and sloped floors are provided for each room or cubicle that has serviceable components containing radioactive liquids. Whenever practicable, drain lines are embedded in concrete floors to provide shielding. If a radioactive drain line must pass through a radiation zone lower than that at which it will terminate, proper shielding is provided. Local gas traps or porous seals are not used for controlled area floor drains. Gas traps are provided at the common sump or tank. Some floor drains are designed to prevent 12.3-6 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) potential backflooding. Whenever practicable, provisions exist to remove plugging in the drain piping. Separate trains of floor drains are laid out for cubicles housing radioactive components from the general personnel corridors and work areas. This minimizes the potential for spreading surface and airborne contamination. The floor drain system layout is discussed in Section 9.3.3. 11. Equipment drains - Drains from equipment carrying radioactive fluids are piped in a hard and closed fashion to the Clean Radwaste (CRW) Drain System. The equipment drainage system is discussed in detail in Section 9.3.3. 12. Radioactive tank vents - Vents are provided for all tanks carrying radioactive waste that are also expected to house radioactive gases. Vents are piped in a closed system to a vent header for collection, filtering, and monitored release of gaseous radioactivity. The radioactive tank vent system is described in Section 11.3. 13. Lighting - Multiple electric lights are provided for each cell or room containing highly radioactive components so that the burnout of a single lamp does not require entry and immediate replacement of the defective lamp. Sufficient illumination remains available. Lighting in a radioactive area is actuated from outside the area, and long-life bulbs are used. Section 9.5.3 describes the lighting system. 14. HVAC - The Heating, Ventilating, and Air Conditioning (HVAC) System is designed to minimize buildup of radioactive contamination and provides easy access and fast replacement of the filter elements. Filter banks and components are separated from adjacent banks and components. Section 12.3.3 provides additional description of the HVAC radiation protection design features. 12.3-7 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 15. Sample stations - Sample stations for routine sampling of process fluids are located in accessible areas. The locations of these sample stations are such that piping and tubing runs are minimized. Shielding is provided at the local sample stations as required to maintain radiation zoning in proximate areas and minimize personnel exposure during sampling. For the unlikely event of a design basis accident (DBA), a post-accident sample station is provided in compliance with NRC regulations set down in NUREG 0737. The station is sufficiently shielded to allow post-accident sampling, sample transport to the onsite facility, and analysis without receiving radiation exposures above 5 rem for the whole body and 75 rem for any extremity. This complies with NRC regulations delineated in NUREG-0737. For a detailed discussion on process and post-accident sampling, see Section 9.3.2. The general layout of the sample stations is shown on Plant Drawings M-23-0 and M-38-0. 16. Clean services - Wherever practicable, clean service components such as compressed air piping, clean water piping, ventilation ducts, and cable trays are not routed through radioactive pipeways. In addition, active components of these clean services are located outside radiation areas wherever possible, to minimize any radiation exposure associated with maintenance of clean system components. 17. Electrical and electronic equipment - Equipment serving electrical or electronic needs is placed in low radiation areas whenever possible. This equipment is non-radioactive, and full access is provided with a minimum of radiation exposures for its surveillance, inspection, maintenance, and repair. 18. Counting room and laboratories - The counting room and radiochemical laboratory facilities are described and discussed in Section 12.5.2. 12.3-8 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) 12.3.1.2 Common Facility and Layout Designs for ALARA This section describes the design features used for standard plant process equipment and layout situations for radioactive systems and for potentially radioactive systems. These features are used in conjunction with the general equipment designs described in Section 12.3.1.1. 1. Valve galleries - Valve galleries are provided with shielded entrances for personnel protection. In many cases, the valve galleries are divided by shielding or distance into subcompartments that service only a few components so that personnel are exposed only to the valves and piping associated with a minimum of components at any given location. Process isolation valves are located close to wall penetrations. Floor drains are provided to collect radioactive leakage. The local areas within the vicinity of the valve gallery areas are provided with decontamination surface coatings. As a minimum, these areas are provided with decontamination surface base (one foot from the floor) and/or decontamination Wainscoat (six feet from the floor). This feature makes it easy to decontaminate surfaces exposed to possible leakage from valves and other components. 2. Piping - Each piping run is analyzed to determine the potential radioactivity level and maximum expected surface dose rate. Radioactive pipes are routed separately from non-radioactive pipes where possible to minimize personnel exposure. Pipes carrying radioactive materials are routed through controlled access areas zoned for a corresponding radiation dose level. Where radioactive piping must be routed through corridors or other low radiation areas, pipes are shielded using special lead or steel details or shielded pipeways. 12.3-9 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) Valves, instruments, or active components are not placed in radioactive pipeways. Wherever practicable, each equipment compartment is used only for those pipes associated with equipment in the compartment. This minimizes exposure due to the operation of one system while maintenance is being performed on another system that is shut down. Piping is designed to minimize low points and dead legs. Drains are provided on piping where low points and dead legs cannot be eliminated. Where possible, thermal expansion loops are raised rather than dropped to prevent crud traps. In radioactive systems, the use of non-removable backing rings in the piping joints is prohibited to eliminate a potential crud trap for radioactive materials. Wherever possible, branch lines having little or no flow during normal operation are connected above the horizontal midplane of the main pipe. Line size changes are typically made by eccentric reducers. Orifices are placed in vertical lines wherever possible. At strategic locations within some selected piping systems, connections are provided for the use of hydrolasers. High pressure hydrolasers will be used in removing pipe surface contaminations and therefore reducing radiation exposures to personnel during maintenance and other operations. Piping carrying resin slurries or evaporator bottoms is run using large radius bends wherever possible, instead of elbows, and horizontal runs are minimized. To prevent possible crud buildup, flow control valves and orifices are used only if they are required for system operation. Large diameter piping is typically used with a minimum number of pipe fittings to reduce crud accumulation. 12.3-10 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 3. Field run piping - Most routing of radioactive piping, large and small, is done by the Bechtel engineering office to ensure that the radiation zone routing is proper and that the aforesaid principles are employed. Where the routing of radioactive piping is delegated to the Bechtel field office, ALARA principles are enforced by way of the ALARA Field Design Guide that was originated by the engineering office. 4. Penetrations - Wherever possible, penetrations are located with an offset between the source and accessible areas to minimize radiation streaming. If offsets are not practicable, penetrations are located as far as possible above the floor elevation to reduce the exposure to personnel. If neither of these two methods is used, then alternative means are employed, such as using baffle shield walls or radiation shielding in the area around the penetration. 5. Contamination control - Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. Equipment vents and drains from radioactive systems are normally piped directly to the radwaste collection system instead of allowing any contaminated fluid to flow across the floor to the floor drain. All welded piping is used for radioactive systems to the maximum extent practicable to reduce system leakage and crud buildup at joints. The valves in some radioactive Nuclear Steam Supply Systems (NSSS) are provided with leakoff connections piped directly to the Radwaste Collection System. They are the main stop, control, bypass, and combined intermediate valves of the main turbine, and the stop and control valves for the reactor feed pump turbines. 12.3-11 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) Decontamination of areas expected to become contaminated within the plant is facilitated by the application of suitable smooth-surfaced coatings to the concrete floors and walls of these areas. Floor drains and sloping floors are provided in all potentially radioactive areas of the plant. In addition, radioactive and potentially radioactive drains are separated from non-radioactive drains. Systems that become highly radioactive, such as the radwaste slurry transport system, are provided with flush and drain connections, for normal or chemical cleaning prior to maintenance. 6. Equipment layout - In systems where process equipment is a major radiation source, e.g., fuel pool and reactor water cleanup, radwaste, condensate demineralizer, etc, the pumps, valves, and instruments are separated from the process component. This allows servicing and maintenance of items in reduced radiation areas. Control panels are located in the lowest dose radiation zones. Redundant equipment is separated from each other, and shielding is provided between the equipment to allow maintenance concurrent with system operation. Major components, e.g., tanks, demineralizers, and filters, in radioactive systems are located in individual shielded compartments. For highly radioactive components, e.g., filters and demineralizers, completely enclosed shielded compartments with hatch openings are provided. Provision is made for some major plant components for removal to lower dose radiation zones for maintenance. Large HVAC filter plenums with multiple filter cartridges are provided with external shielding. 12.3-12 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) Labyrinth entryway shields or shielding doors are provided for each compartment from which radiation could shine to access areas and exceed the radiation dose limits for those areas. Adequate space for removal of components is provided. Plant Drawings N-1031, N-1033, N-1046 and N-1012 provide typical layout arrangements for demineralizers, liquid and gaseous radwaste filters, waste sludge tanks, radwaste evaporators, off-gas recombiners, sample stations, charcoal beds, and their associated valve compartments or galleries. 7. Special provisions for surveillance of radioactive system areas - To minimize plant personnel radiation exposure during routine area surveillances of radioactive systems, several special design features are provided. A number of radiation shielding viewing windows are installed at strategic locations in the shield walls surrounding radiation areas. Such areas as the pump and valve galleries of the radwaste collection and processing systems, the pump and heat exchanger rooms of the Reactor Water Cleanup (RWCU) and Fuel Pool Cooling and Cleanup (FPCC) Systems, and the resin regeneration station are provided with viewing windows. These windows will allow surveillance of active system components from the personnel corridors without the need to enter the radiation areas. For special viewing and remote control purposes, television cameras are provided in areas such as the radwaste drum filling turntable stations, drum conveyor aisle, and the drum storage hayloft, where high radiation fields can be expected, surveillance of the area is necessary, and the presence of individuals is prohibited. 12.3-13 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) Radioactive and non-radioactive systems are normally separated to limit radiation exposure from routine inspection of non-radioactive systems. For radioactive systems, emphasis is placed on adequate space and ease of motion in a shielded inspection area. Where longer times for routine inspection are required and permanent shielding is not feasible, sufficient space for portable shielding is normally provided. In high radiation areas where routine surveillance is required, remote viewing devices are provided as needed. Typically, equipment and valves in high radiation areas are made easily accessible by providing permanent access platforms, easily removable insulation, etc. Equipment manways are readily accessible. Equipment laydown area requirements are considered in the layout, and adequate space is provided where necessary. 8. Facilities for handling unsealed radioactive material - As discussed in Section 12.2.1.7, special materials used in the radiochemistry laboratory require the design of special storage and handling equipment. For the handling of radioactive unsealed materials, the following are provided: a. Exhaust hoods that exhaust to the ventilation system are located in areas such as sample stations and the radiochemistry laboratory. b. Sample sinks for the collection and control of radioactive liquids at sample stations and the radiochemistry laboratory. c. Decontamination facilities, the radiochemistry laboratory, restricted machine shop, and instrument calibration and repair shop are situated at various locations in the plant and are described in Section 12.5. 12.3-14 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
- d. An area for the repair and maintenance of removed control rod drives (CRDs) is provided in the reactor enclosure near the
CRD removal hatch.
- e. Area and airborne exhaust duct radiation monitor systems alert personnel to any abnormal condition due to the handling
of unsealed sources.
12.3.1.3 Radiation Zoning and Access Control Access to areas inside the plant structures and plant yard is regulated and controlled. Each radiation area, as defined in 10CFR20, is provided with a personnel alert sign, each high radiation area (10CFR20) with an expected dose rate between 100 mrem/h and 1 rem/h will be conspicuously posted "CAUTION, HIGH and controlled with an RWP while high radiation areas with an expected dose rate greater than 1 rem/h is provided with a locked personnel barrier, as shown on Plant Drawings N-1031 through N-1038, N-1041 through N-1047 and N-1011 through N-1016 and described in Section 12.3.2.2.11. Very high radiation areas, as defined in 10CFR20, are conspicuously posted "GRAVE DANGER, VERY HIGH RADIATION AREA" and provided with a locked personnel barrier. Section 12.5 describes both the control of ingress and egress of plant operating personnel to radiologically controlled access areas, and the procedures employed to ensure that personnel exposure is within the limits prescribed by 10CFR20.
All plant areas are categorized into radiation zones according to expected radiation levels. Each radiation zone defines either the highest component dose rate in the area or the radiation level from all contributing sources inside and outside of the area, whichever is higher. Each room, corridor, and pipeway of every plant structure is evaluated for potential radiation sources during normal operation, including anticipated operational occurrences and shutdown. The radiation zone categories used, and their descriptions, are given in Table 12.3-1, and the specific zoning for each plant area during normal operation and plant cold shutdown is shown on Plant Drawings N-1031 through N-1038, N-1041 through N-1047 and N-1011 through N-1016 and Figures 12.3-22 through 12.3
-29, and described in Section 12.3.2.2.11. All frequently used areas, such as corridors are shielded for Zone I or Zone II access. 12.3-15 HCGS-UFSAR Revision 22 May 9, 2017
The locations of airborne radioactivity and area radiation monitors are described in Section 12.3.4. As a special radiation protection design feature, plant personnel are able to establish temporary control points. At the entrances to radioactive areas housing equipment expected to need routine maintenance and repair, sufficient space is allocated for placing an approximately 100
-ft 2 contamination control area and a step-off pad. In addition to several 55-gallon drums to hold contaminated clothing, a table and chair can be moved in for radiation protection personnel controlling the ingress/egress of personnel working in the radioactive room. These temporary
control points control and minimize any spread of contamination.
Section 12.3.2 provides a discussion of post
-accident radiation zoning and shielding for vital plant areas, and post
-accident emergency activities.
(Historical Information) 12.3.1.4 Control of Activated Corrosion Products
To minimize the radiation exposure associated with the deposit of activated corrosion products in reactor coolant and auxiliary systems, the following
steps have been taken:
- 1. The Reactor Coolant System (RCS) consists mainly of austenitic stainless steel, carbon steel, and low alloy steel components. The nickel metal content of these materials is low, in accordance with applicable ASME material specifications, because nickel metal contains trace amounts of cobalt, and cobalt 60 is one of the predominant radioactive corrosion products.
12.3-16 HCGS-UFSAR Revision 21 November 9, 2015
(Historical Information) 2. Powdex filter demineralizer filtration is employed in the use of the RWCU system. This system is in continuous operation for the cleanup of the reactor water during full power operation, as well as during plant shutdowns. The system is described in Section 5.4.8. This filtration removes activated corrosion products and crud, and helps to keep down the radioactivity level of the reactor coolant. 3. Valve packing materials are selected primarily for their properties for use in the particular nuclear environments. 4. Various flushing, hydrolasing, draining, testing, and chemical addition connections have been incorporated into the design of piping and equipment that handle radioactive materials. These connections are used if corrosion product removal is to be performed. 5. The plant is designed with a full flow, deep bed condensate demineralizer system for the feedwater. This filtration system helps to minimize the buildup of corrosion products and other impurities in the Reactor Coolant System. This system is described in Section 10.4.6. 6. The Condensate Pre-filter system has been added to the plant upstream of the deep bed condensate demineralizers. This system operates at 100% condensate flow to remove insoluble impurities, primarily iron, in order to improve the performance of the deep bed demineralizers and to reduce corrosion products in the Reactor Coolant system. This system is described in Section 10.4.6.6. 12.3.2 Shielding In this section, the bases for the nuclear radiation shielding design and the shielding layout are discussed. Also discussed are the mathematical methods and geometric models used in the calculations for the radiation shielding design. 12.3-17 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 12.3.2.1 Design Objectives The basic objective of the plant nuclear radiation shielding is to reduce personnel exposures. In conjunction with a program of controlled personnel access to and occupancy of radiation areas, radiation levels are held within the allowable limits of NRC Regulations 10CFR20 and 10CFR50, and are as low as is reasonably achievable (ALARA), in accordance with Regulatory Guide 8.8. Radiation shielding, as well as the layout and the design of equipment, components, and piping, are considered in ensuring that exposures are kept ALARA during anticipated personnel activities in areas of the plant expected to contain radioactive materials. Basic plant conditions considered in the radiation shielding design are normal operation at full power, plant shutdowns, and anticipated operational occurrences. The shielding design objectives for the basic plant conditions are: 1. To ensure that radiation exposures to plant operating personnel, contractors, administrators, visitors, and proximate site boundary occupants are ALARA and within the limits of 10CFR20. 2. To ensure sufficient personnel access and occupancy time for normal anticipated surveillance, inspection, maintenance, repair, and safety-related test operations as required for each plant equipment, valve, instrumentation, and piping area. 3. To reduce potential equipment neutron activation and minimize the possibility of radiation damage to materials and components from neutron, gamma, and beta exposures. For the unlikely event of a design basis accident (DBA), including a loss-of-coolant accident (LOCA), radiation shielding is provided for vital areas, such as the plant control room, to keep exposures within the guideline values of 10CFR50.67, and also to protect the general public and ensure that their accident exposures remain within the guideline values of limits of 10CFR50.67. For further discussion of the protection of plant vital areas, see Section 12.3.2.2.6. 12.3-18 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 12.3.2.2 General Radiation Shielding Design For the protection of individuals, radiation shielding structures are provided to attenuate direct and scattered radiation to less than the upper limits of the radiation zones for each plant area, as shown on the radiation shielding and zoning drawings. For the definitions of radiation zones and personnel radiological controls, see Plant Drawings N-1031 through N-1038, N-1041 through N-1047 and N-1011 through N-1016, Figures 12.3-22 through 12.3-28 and Section 12.3.2.2.11. The minimum shielding requirements for all plant areas are shown on those shielding and radiation zoning drawings. General arrangement of the plant areas, structures, and equipment discussed in this section are also shown on the radiation shielding and zoning drawings. The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 147 lb/ft3. Wherever poured in place concrete is replaced by concrete blocks or other material, the design ensures protection on an equivalent shielding basis, as determined by the radiation attenuation characteristics of the concrete block or the other material selected. High density concrete having a minimum density of 200 lb/ft3 is used where space limitations do not allow use of ordinary concrete. In a few cases, special shielding design details use steel and/or lead materials where space limitations or layout complexities make the use of concrete impossible or uneconomical. For a number of entrances to highly radioactive areas, heavy solid steel shield doors are provided. This type of shielding is used only when other means of protection, such as labyrinths, cannot be employed. Water is also used as a primary shield material for areas surrounding the spent fuel transfer and storage areas. 12.3-19 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) Special features incorporated in the structural layout and design of the plant shielding are introduced to maintain radiation exposures ALARA in routinely occupied areas, such as valve operating stations and pump rooms. These are described in more detail under Radiation Protection Design Features in Section 12.3.1. As discussed in Section 1.8, Regulatory Guide 1.69 is used for Hope Creek Generating Station (HCGS) in the design of nuclear radiation shielding. Regulatory Guide 8.8 is also used. Concrete shielding properties and standards are also discussed in Section 3.8. The shielding thicknesses are selected to reduce the aggregate computed radiation level from all contributing sources below the upper limit of the radiation zone specified for each plant area. Shielding requirements are evaluated at the point of maximum radiation dose through any wall or floor slab. Therefore, the actual anticipated radiation levels in the greater region of each plant area are less than this maximum dose and are also less than the radiation zone upper limit. The labyrinths are constructed such that the scattered dose rate, plus the directly transmitted dose rate through the shield wall from all contributing sources, is below the upper limit of the radiation zone specified for the particular plant area exposed. 12.3.2.2.1 Drywell and Reactor Building Shielding Design During reactor operation and shutdown, the reinforced concrete drywell shield wall and the Reactor Building walls protect both personnel occupying adjacent plant structures, and yard areas, from radiation originating in the reactor vessel and associated equipment. Where personnel or equipment hatches or system penetrations pass through the drywell shield wall, additional shielding is designed to attenuate the radiation to below the required level, as defined by the radiation zones outside the drywell wall during normal full power operation and shutdown. The drywell and Reactor Building shielding also functions as protection during accidents, including LOCAs, and ensures that exposure levels in the vital plant areas remain within limits as defined by 10CFR5067. 12.3-20 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 12.3.2.2.2 Drywell Interior Shielding Design During plant operation, areas within the drywell are designed for Zone VI and higher levels, and they are inaccessible (See Plant Drawing N-1043). The biological shield, which surrounds the reactor pressure vessel (RPV), provides shielding for access in the drywell during shutdown. It also reduces the activation of, and radiation damage to, drywell equipment and materials during full power operations. Special shield doors are provided for a number of pipe penetrations through the biological shield. When inservice inspections are required during plant outages, these doors allow access to specific locations at the reactor pressure vessel and its pipe nozzles. 12.3.2.2.3 Reactor Building Interior Shielding Design The drywell shield wall concrete shell structure is designed to reduce radiation levels in normally occupied areas of the Reactor Building, from sources such as the RPV and the main steam lines within the drywell, to less than the maximum level for a Zone II. As defined on Plant Drawings N-1031 through N-1038, N-1041 through N-1047 and N-1011 through N-1016, and Figures 12.3-22 through 12.3-28, this zone level allows fulltime access and occupancy of personnel corridor and general work areas located outside the drywell and inside the reactor building. Penetrations and hatch openings in the drywell concrete shield wall are shielded, as necessary, to meet adjacent area radiation zoning levels. Shielding requirements for the personnel, equipment, and control rod drive (CRD) removal hatch openings are shown on Plant Drawing N-1043 for the areas numbered 4330, 4322, and 4326, respectively. Drywell piping and electrical penetrations are shielded by providing either local shields within the penetration assembly or a shielded pipe chase. Shielded pipe chase locations and bulk shielding requirements are shown on Plant Drawings N-1043 through N-1046. 12.3-21 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) These pipe chases, with room numbers 4316, 4319, 4321, 4327, 4329, 4402, 4505, and 4509 are designated as radiation areas having Zone III radiation levels and higher during reactor power operation and are provided with personnel access control barriers. Adjacent to certain vent lines below Elevation 100 feet-0 inches, the shield wall grout fill has been excavated to create tunnels without significant shielding. Room 4102, which is adjacent to the shield wall in these areas, is designated as having Zone VII radiation levels during reactor power operations and is provided with personnel access control barriers. The components of the Reactor Water Cleanup (RWCU) System, most of which are in the Reactor Building and are described in Section 5.4, are located in shielded compartments. The RWCU compartments are designed as areas having Zone V radiation levels and higher and are restricted access areas. Shielding is provided for each piece of equipment in the RWCU system consistent with its calculated maximum activity, as described in Section 12.2, and with the access and zoning requirements of the adjacent areas. This equipment includes: 1. Regenerative heat exchanger 2. Nonregenerative heat exchanger 3. RWCU recirculation pumps and piping 4. RWCU filter demineralizers and precoat and holding pumps 5. RWCU backwash receiving tanks, transfer pumps, and piping. The Residual Heat Removal (RHR) System pumps and heat exchangers are in operation after reactor shutdown to remove reactor core decay heat. The radiation levels in the vicinity of this equipment may temporarily reach as high as Zone VI levels due to activation and fission products in the reactor water. Shielding is designed to attenuate radiation from RHR equipment during shutdown cooling operations to levels consistent with the radiation zoning requirements of adjacent areas. 12.3-22 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) The Reactor Building also houses the Reactor Core Isolation Cooling (RCIC) System and the High Pressure Coolant Injection (HPCI) System. These systems require periodic functional tests. They use the radioactive main steam for the pump turbine drive. Shielding protects the surrounding areas according to applicable access requirements. Since HCGS does not have a cattle chute design shield, actions will be taken such as administrative controls, remote monitoring, and physical barriers to preclude access to the drywell during fuel movement. However, personnel may be permitted limited access to the lower levels for necessary work during refueling operations. One of the primary radiation sources inside the reactor building is the spent fuel assemblies. Spent fuel transfer and storage is performed underwater in the fuel transfer canal and in the spent fuel storage pool. Water and concrete shielding are provided for areas surrounding the fuel transfer canal and the storage pool, to ensure that radiation levels remain below zone levels specified for adjacent areas. A portable, shielded fuel transfer chute also is installed in the reactor cavity during refueling operations to provide additional shielding to upper drywell areas. Water is also used as shielding material for the steam dryer and separator storage pool. Concrete walls surrounding the pool and water in the pool are designed to provide Zone II dose rates in adjacent accessible areas during storage of the dryer and separator. The Fuel Pool Cooling and Cleanup (FPCC) System shielding is based on the maximum activity discussed in Section 12.2 and the access and zoning requirements for adjacent areas. Equipment in the FPCC system to be shielded includes the FPCC heat exchangers, fuel pool filter/demineralizers, pumps, and piping. The spent fuel shipping cask loading pit is designed and shielded for the safe transfer of spent fuel assemblies from the storage pool into the cask. The assemblies are transferred underwater, in order to protect surrounding floor areas to Zone II levels. 12.3-23 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) The traversing in-core gamma probe (TIP) system drive and housing components are located inside a shielded compartment to protect personnel from the neutron activated portions of the TIP cables. Concrete shielding is provided for the section of the main steam tunnel that is located within the Reactor Building. The high energy gamma radiation emanating from the main steam is reduced by shielding to allow full personnel access in adjacent areas. For the CRD removal, repair and decontamination area shielding is provided to protect the adjacent personnel work and corridor areas. Special lead shielding caps are provided on the CRD spuds. 12.3.2.2.4 Radwaste Areas Shielding Design In the radwaste areas where gaseous, liquid, and solid radioactive wastes are collected, stored, and processed, shielding is provided to ensure that the radiation zoning and access requirements are met for areas occupied by plant personnel. The following major equipment, which contains and handles liquid radwaste, is placed in individual shielded cubicles, as shown on Plant Drawings N-1031, N-1032, N-1037, N-1033 and N-1034: 1. Decontamination solution concentrated waste tank 2. Radwaste regeneration vessel 3. Radwaste resin holding tank 4. Concentrated waste tanks 5. Waste neutralizer tanks 6. Waste surge tank 12.3-24 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) 7. Floor drain sample tanks 8. Waste sample tanks 9. Waste collector tanks 10. Cleanup phase separators 11. Spent resin tank 12. Waste sludge phase separator 13. Chemical waste tank 14. Decontamination solution concentrator 15. Floor drain collector tanks 16. Detergent drain tank 17. Waste evaporators 18. Fuel pool filter/demineralizers 19. Waste and floor drain filters 20. Waste and floor drain demineralizers. The pumps, valve manifolds and instrumentation, and general piping associated with the equipment listed above are also shielded in cubicles, galleries, and pipeways, respectively. The following major equipment, which contains and handles gaseous radwaste, is placed in individual shielded cubicles, as shown on Plant Drawings N-1031 and N-1032: 12.3-25 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) 1. Feed gas cooler condenser 2. Preheater 3. Recombiner 4. Offgas holdup piping 5. Passive glycol cooler condenser 6. Charcoal guard bed 7. Charcoal adsorber vessels 8. High efficiency particulate air/absolute (HEPA) filters. Valves, instruments, and general piping associated with the aforesaid equipment are also shielded in cubicles, galleries, and pipeways. For the solidification, drumming, storing, and shipping of liquid and solid radioactive wastes, the following major equipment is placed in shielded individual cubicles, as shown on Plant Drawings N-1033, N-1034 and N-1035: 1. Centrifuge feed tank 2. Waste centrifuge 3. Vapor body 4. Extruder evaporators 5. Entrainment separator 6. Concentrated bottoms tank 12.3-26 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) 7. Solid waste compactor 8. Drum capping, swiping, and conveyor equipment. The storage and shipping areas for the solidified waste drums are heavily shielded. The pumps, valves and instruments, and general piping associated with the major equipment listed above, are also shielded in cubicles, galleries, and pipe chases, respectively. 12.3.2.2.5 Turbine Building Shielding Design Radiation shielding is placed around the following major equipment inside the Turbine Building in order to comply with zone access requirements, shown on Plant Drawings N-1011 through N-1016, and exposure restrictions for adjacent areas: 1. Turbine condensers 2. Turbine hotwells 3. Primary condensate pumps 4. Steam jet air ejectors 5. Off-gas feed piping 6. Steam packing exhauster condenser 7. Seal water coolers 8. Mechanical vacuum pump 9. Condensate demineralizers 10. Cation and anion regeneration vessels 12.3-27 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information) 11. Resin mix and hold tank 12. Ultrasonic cleaning equipment 13. Feedwater heaters 14. Reactor feed pump turbines 15. Steam seal evaporators 16. Moisture separators 17. High and low pressure turbines 18. Main steam piping. 19. Backwash receiving tank, tipped HIC and transfer pumps of the condensate pre-filter system Shielding is also provided at the turbine building operating floor because of the relatively close location of Zone I areas. These areas are offices, technical document and training facilities, quality engineering (QE) and conference rooms, lobbies, and toilet facilities that are designed for unlimited personnel access. 12.3.2.2.6 Post Accident Shielding Design and Access Review A post accident shielding and access review was performed to ensure the accessibility of vital areas in which personnel will or may be present to perform mitigation or monitoring functions during post accident operations. 1. Source terms - Following a postulated accident where substantial core damage has occurred, radioactive materials will be released from the fuel due to fuel rod cladding failure. 12.3-28 HCGS-UFSAR Revision 16 May 15, 2008
- 2. Not Used 3. Airborne sources in the Auxiliary and Turbine Buildings - The transport pathway of the airborne sources in the Auxiliary Building (control and diesel generator areas, and service areas), and Turbine Building consists of leakage: a. from the drywell to the reactor building, and discharge to the environment through the Filtration, Recirculation, and Ventilation System (FRVS) b. from engineered safety feature components outside the primary containment, and discharge to the environment through the FRVS c. from the main steam isolation valves (MSIVs), and discharge to the environment from the Turbine Building The airborne activity then re-enters the buildings through the ventilation intake systems after dilution in the atmosphere. The fission product release to the environment is described in Section 15.6.5.5. The atmospheric dispersion factors for the airborne transport pathways to the control room emergency intake are given in Table 6.4-2. The atmospheric dispersion factors used in evaluating the total integrated doses at all other locations within the building wake cavity are given in Table 12.3-10. 12.3-29 HCGS-UFSAR Revision 18 May 10, 2011
- 4. Not Used 12.3-30 HCGS-UFSAR Revision 12 May 3, 2002
(Historical Information)
12.3.2.2.7 Diesel Generator Areas Shielding Design The diesel generator areas are part of the control and diesel generator complex of the Auxiliary Building. There are no radiation sources in the diesel generator areas. Therefore, no shielding is required for these areas from internal sources. For post accident access, these areas have sufficient
shielding to be protected from drywell and Reactor Building radiation shine.
12.3.2.2.8 Deleted
12.3-32 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information) 12.3.2.2.9 Counting Room Shielding Design The plant counting room, which is part of the radiochemistry laboratory facility in the Auxiliary Building, has received special attention with regard to radiation shielding. Because the counting room contains instruments very sensitive to radioactivity, it is imperative that the background radiation levels be minimized. To accomplish this, no flash is used in the concrete mix for the walls and slabs surrounding the counting room. Flash normally contains a relatively large amount of slow decaying natural radioactive isotopes. In addition, the shield walls and slabs are sized to maintain a background radiation level of less than 130 mrem/year for anticipated operational occurrences and 45 mrem/year for normal operation. 12.3.2.2.10 General Plant Yard Areas Shielding Design All general plant yard areas are at Zone I radiation levels, and are, therefore, fully accessible during all phases of normal full power operation. Only the condensate storage tank (CST) area is an exception, having a Zone II designation. This area is surrounded by a wall shielding the adjacent Zone I plant yard areas. 12.3.2.2.11 Low Level Radwaste Storage Facility Shielding Design Shielding is provided to ensure that the radiation zoning and access requirements are met for areas occupied by plant personnel. Shielding is designed to provide Zone I dose rates outside the facility when DAW boxes are being moved and/or when no crane operations are taking place. When a HIC or Pallet is being moved with the crane, dose rates outside the facility will fall within Zones III, IV or V depending on the contact dose rate on the containers. The DAW storage area is designated Zone IV and the vault storage area is designated Zone VIII. The control room is designated Zone II. The truck bay is normally Zone II or III but will be higher when waste is being moved. During periods when offsite disposal or storage is available and radwaste is not stored in the facility, zone designations will be assigned as appropriate for any alternate use of the facility. (Historical Information) 12.3.2.3 Shielding Calculational Methods and Geometry Models The shielding design of the structures provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure in accordance with ALARA is based on maximum equipment activities under the plant operating conditions described in Section 12.2. The thickness of each shield wall surrounding radioactive equipment is determined by approximating the actual geometry and physical conditions of the source or sources as closely as possible. The isotopic concentrations are converted to energy group sources using data from standard 12.3-33 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) references that are accepted and widely used by the nuclear industry. This is from material derived from References 12.3-3 through 12.3-5. The geometric model assumed for the shielding evaluation of tanks, heat exchangers, filters, demineralizers, and evaporators is a finite cylindrical volume source. For the shielding evaluation of piping, the geometric model is a finite or infinite shielded cylinder. In cases where radioactive materials are deposited on cylindrical component surfaces, e.g., pipes, this is treated as an annular cylindrical surface source. Slab source, truncated cone, or flat surface source geometries, either finite or infinite, are employed in shielding analyses for extended flat face radiation sources, such as are encountered with the box type storage arrangement of spent fuel assemblies. Point source geometry is employed for highly concentrated sources having relatively small dimensions. Examples are component crud hot spots or instrument calibration sources. Sources having large dimensions are encountered in the analyses of post accident fission product gas clouds. The analyses in this case use infinite cylindrical or spherical source geometries. Typical computer codes that are used for shielding analysis are listed in Table 12.3-4. Shielding attenuation data used in these codes include gamma attenuation coefficients, found in Reference 12.3-16; gamma buildup factors, found in Reference 12.3-17; neutron-gamma multigroup cross sections, found in Reference 12.3-18; and albedos, mentioned in Reference 12.3-19. Where shielded entryways to compartments containing high radiation sources are necessary, labyrinths or mazes are designed using a general purpose gamma-ray scattering Code G-33, mentioned in Reference 12.3-15. 12.3-34 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 12.3.3 Ventilation The plant ventilation systems are designed to provide a suitable environment for personnel and equipment during normal plant operation and anticipated abnormal occurrences. Detailed Heating, Ventilating, and Air Conditioning (HVAC) System descriptions, including compliance with and exceptions to Regulatory Guides 1.52 and 1.97, are provided in Sections 1.8 and 9.4. Control room habitability is discussed in Section 6.4. Instrumentation and controls for Engineered Safety Feature (ESF) HVAC Systems are discussed in Section 7.3. Section 7.4 discusses the safe shutdown HVAC system. 12.3.3.1 Design Objectives The ventilation systems are designed to maintain inplant airborne activity levels within the limits of 10CFR20 in personnel access areas and also operate to prevent the spread of airborne radioactivity during normal plant operation and anticipated abnormal occurrences. During post accident conditions, the ventilation systems for the main control room (MCR) and technical support center (TSC) help to provide a suitable environment for personnel and equipment to ensure continuous occupancy in these areas, except for a fire/smoke condition, when personnel are to be evacuated. If smoke is detected in the outside air intake duct to the MCR or TSC, the affected system can be manually switched over to the recirculation mode. The plant ventilation systems are designed to comply with the airborne radioactivity release limits for offsite areas during normal and abnormal plant operations. 12.3.3.2 Design Criteria Design criteria for the plant ventilation systems include the following: 12.3-35 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 1. During normal operation and anticipated abnormal occurrences, the average and maximum airborne radioactivity levels to which plant personnel are exposed in restricted areas of the plant are as low as is reasonably achievable (ALARA) and within the limits specified in 10CFR20. The average and maximum airborne radioactivity levels in unrestricted areas of the plant during normal operation and anticipated abnormal occurrences are ALARA and within the limits of 10CFR20. 2. During normal operation and anticipated abnormal operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary are ALARA and within the limits specified in 10CFR20 and 10CFR50. 3. The plant site dose guidelines of 10CFR100 are satisfied following those hypothetical accidents, described in Section 15, which involve a release of radioactivity from the plant. 4. The dose to MCR and TSC personnel must not exceed the limits specified in GDC 19 of Appendix A to 10CFR50, following those hypothetical accidents, described in Section 15, which involve a release of radioactivity from the plant. 12.3.3.3 Design Guidelines 12.3.3.3.1 Guidelines to Minimize Airborne Radioactivity To accomplish the design objectives, the following guidelines are used in plant design to minimize airborne radioactivity: 12.3-36 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 1. Access control and traffic patterns are considered in the basic plant layout to minimize the exposure to contamination. 2. Radioactive vents and drains from equipment and piping are piped directly to the collection system, instead of allowing contaminated fluid to flow openly across the floor to the floor drain. 3. All welded piping is used on contaminated systems to reduce system leakage. See Section 11.1.5 for additional discussion. 4. Metal diaphragm or bellows seal valves are used in those radioactive piping systems where no leakage can be tolerated. 5. Coatings are applied to the concrete floors and walls of potentially contaminated areas to facilitate any decontamination. 6. Potentially contaminated equipment has design features that minimize the potential for airborne radioactive contamination during maintenance operations. These features include flush connections on pump casings, for draining and flushing the pumps prior to maintenance, and flush connections on piping systems that could be highly radioactive. 7. Exhaust ventilation is used in plant areas to direct any airborne radioactive contaminants away from the personnel breathing areas and into the ventilation filtering systems. High efficiency air filters are used in all supply air systems to minimize particulate inventory within the plant. 12.3-37 HCGS-UFSAR Revision 16 May 15, 2008
(Historical Information) During refueling operations in the reactor building, at least two of three supply and exhaust fans (three 50 percent capacity supply and three 50 percent capacity exhaust) are operated, and manual duct dampers are repositioned to provide increased ventilation capacity. 6. In the radwaste area systems, the supply and exhaust fans are interlocked to ensure exhaust fan operation prior to supply fan operation. 7. Individual areas can be isolated upon indication of contamination to prevent the discharge of contaminants to the environment. 8. Redundant containment isolation dampers are installed at ventilation duct penetrations in accordance with 10CFR50, Appendix A, GDC 54 and 56, including valve controls, to ensure that the containment integrity is maintained. See Section 6.2.4 for additional discussion. 9. Tornado dampers are provided in outside air intake and exhaust ducts in safety-related HVAC systems, where it is necessary to protect HVAC equipment, ductwork, and process equipment in the event of a tornado. 10. The Reactor Building Ventilation System (RBVS) supply air quantity is adjusted so that a negative building pressure is maintained during normal operation. 11. Before the drywell and suppression chamber are purged by the RBVS, the containment atmosphere is recirculated through the Containment Prepurge Cleanup System (CPCS) to reduce atmospheric radioactivity as required. 12.3-39 HCGS-UFSAR Revision 16 May 15, 2008
(Historical Information) 5. HEPA filters are designed and tested to be free of bypass leakage when clamped in place against compression seals. Each filter housing is designed and tested to be airtight with bulkhead doors that are closed against compression seals. 6. Potentially contaminated filter equipment is shielded to reduce exposure to personnel and equipment in surrounding areas. 7. Particulate filters are changed when required by the indicated pressure drop across the filter bank. Charcoal adsorbers are changed when required by the residual adsorption capacity of the bed. Removable test canisters are located in the carbon bed. The testing of the carbon adsorbers and all other components is described in Sections 6.5.1 and 9.4. 12.3.3.4 Design Description The ventilation systems serving the Reactor Building, Auxiliary Building radwaste area, and Turbine Building are assumed to be potentially radioactive, and are discussed in detail in Section 9.4. Although the MCR and TSC are considered to be non-radioactive areas, radiation protection is provided to ensure habitability, as described in Section 6.4. Ventilation system design parameters are given in Tables 12.3-5 through 12.3-8. 12.3-41 HCGS-UFSAR Revision 16 May 15, 2008 Typical shielding arrangement for the potentially radioactive charcoal and HEPA filter units is shown on Figure 12.3-58. 12.3.4 Area Radiation and Airborne Radioactive Materials Monitoring Instrumentation 12.3.4.1 Area Radiation Monitoring Design 12.3.4.1.1 Design Bases The Area Radiation Monitoring System (ARMS) is provided to supplement the personnel and area radiation survey provisions of the plant radiation protection program, described in Section 12.5, to ensure compliance with the personnel radiation protection guidelines of 10CFR20, 10CFR50, 10CFR70, and Regulatory Guides 8.2, 8.8, and 8.10. Consistent with this purpose, the area radiation monitors function to: 1. Immediately alert plant personnel entering or working in background or low radiation areas of increasing or abnormally high radiation fluxes that could result in inadvertent exposures if unnoticed. 2. Inform the control room operator of the occurrence and approximate location of an abnormal radiation flux increase in background or low radiation areas, and provide a continuous record for operating and health radiation protection. 3. Comply with the requirements of 10CFR50, Appendix A, GDC 63, for monitoring fuel and waste storage and handling areas. 4. Alarm by means of three area monitors, which are located near the new and spent fuel storage areas. 12.3-42 HCGS-UFSAR Revision 0 April 11, 1988
- 5. Assist, in general, in maintaining personnel exposures as low as reasonably achievable (ALARA). The ARMS has no function related to the safe shutdown of the plant, or to the quantitative monitoring of releases of radioactive material to the environment. Training and qualification of radiation protection, and other personnel assigned to use and maintain the ARMS, follow ANSI/ANS-3.1-1981 guidance. 12.3.4.1.2 Criteria for Area Monitor Selection The following design criteria are applicable to the Area Radiation Monitoring System: 1. Rangeability - There are five decades of range with the high alarm setpoint preferably not lower than the second decade, and not higher than the fourth decade, and set at the maximum allowable exposure rate for the area being monitored. The lower range limit is either natural background or one decade below the normal operating level of each particular area. 2. Overrange Response - The system continues to read off-range upscale if exposed to radiation levels above maximum range. 3. Sensitivity - The system is sensitive to gamma energies of 80 keV and above. 4. Response - In any range, the readout indicates at least 90 percent of its endpoint reading within 5 seconds after a step change in radiation flux at the detector. 5. Energy Dependence - The indicated exposure rate mrem/h at the local indicating unit (LIU) is within 20 percent of 12.3-43 HCGS-UFSAR Revision 0 April 11, 1988 the actual dose rate in each detected area for gamma energies between 80 keV and 2.5 MeV. 6. Environmental Dependence - The system meets the above requirements for all variations of temperature, pressure, and relative humidity within each area delineated in Section 3.11. 7. Exposure Life - Each detector maintains its characteristics up to an integrated exposure of 105 rads. 8. Electromagnetic Interference (EMI) - EMI from any source in the operating environment will not disturb the meter indication at the (LIU). 9. Accuracy - The value of radiation flux indicated in the main control room will be ~ 2 percent of the value indicated at the LIU. 10. Drift - The drift of any detector indication during one year will be no more than 2 percent of the actual value in the midranges. 12.3.4.1.3 Criteria for Location of Area Monitors Area radiation monitors are provided in areas to which personnel normally have access, and in which there is a potential for personnel to receive radiation doses in excess of 10CFR20 limits in a short period of time because of system failure or improper personnel action. The detectors are wall mounted and are located so that the flux measurement is as representative as possible of the area. The detectors are designed and manufactured to be suitable for their locations. Any plant area that meets one or more of the following criteria is monitored: 1. Zone II areas where personnel could otherwise unknowingly receive high levels of radiation exposure because of process system failure or personnel error 12.3-44 HCGS-UFSAR Revision 10 September 30, 1999
- 2. New fuel storage: two detectors are installed to provide area radiation monitoring 3. At HCGS, area monitors are provided in accordance with GDC 63 of 10CFR50 Appendix A. Acceptance Criteria II.B.17 of standard review plan 12.3 - 12.4 provides criteria for the establishment of locations for fixed continuous area gamma radiation monitors. The specific document referenced is ANSI/ANS-HPSSC-6.8.1-1981. The locations and numbers of monitors used at HCGS are not in full compliance with this standard. The location of these monitors is in the vicinity of personnel access areas only. These locations are based on the dose assessment and operating experiences from other nuclear power plants. In addition, these locations were finalized prior to the issuance of this standard and provide an acceptable method of monitoring area radiation levels. Acceptance Criterion II.4.b.3 requires ventilation monitors to be placed upstream of the high efficiency particulate air (HEPA) filters. Discussion on the ventilation monitors is provided in Section 12.3.4.2.2. Acceptance Criterion II.4.a.3, of the standard review plan 12.3 -12.4, provides criteria for on-scale readings of dose rates for normal and anticipated operational occurrences and accidents. The on-scale reading ranges designed for the monitors at HCGS do not in all cases comply with this standard. The general area and airborne ventilation monitors do not have post accident functions. Only a few selected monitors, located in vital areas such as the plant control room and the technical support center, have post accident functions and are designed in compliance. (Also see Section 11.5.) Acceptance Criteria II.4.a.8 and II.4.b.7 of the standard review plan 12.3 - 12.4 provide criteria for emergency power. The HCGS 12.3-45 HCGS-UFSAR Revision 10 September 30, 1999 design is not in compliance with the standards. None of these monitors in question have any safety-related functions and are not on emergency power. Acceptance Criteria II.4.e of Standard Review Plan 12.3-12.4 identifies the requirements to provide instrumentation capable of monitoring accidental criticality in accordance with the requirements of 10CFR70.24(a)(1), Regulatory Guide 8.12 and ANSI Standard N16.2. Based upon NRC evaluation of the information presented in the Hope Creek Special Nuclear Material (SNM) License Application dated May 23, 1985, PSE&G has been granted exemption to 10CFR70.24 as documented in the Hope Creek SNM License No. 1953 dated August 21, 1985. When the Special Nuclear Material License expired, the exemption conditions were incorporated into the Operating License in SSER 5. These conditions are specific to GE fuel only. Alternately, 10CFR50.68 can be used to demonstrate compliance with 10CFR70.24. both of these approaches eliminate the need for the instrumentation since criticality is not credible. 12.3.4.1.4 System Description (Area Radiation Monitoring) The Area Radiation Monitoring System detects, measures, and indicates ambient gamma radiation fluxes at various locations in the plant. It also provides audible and visual alarms in areas monitored, and in the main control room if the gamma radiation exceeds a specified limit. Local indicator units (LIUs) indicate the gamma flux in the area monitored. An indication is provided by a main control room annunciator if there is an alarm or a malfunction in any area monitor. Each area radiation monitoring channel consists of a detector and LIU that provide radiation flux indication and alarm at or near the detector location. Radiation indication and trip status are indicated on the CRT of the display/keyboard/ printer (DKP) and indicated on the CRIDs and SPDS CRTs located in the main control room. The 12.3-46 HCGS-UFSAR Revision 15 October 27, 2006 area radiation monitor provided in the main control room has no local alarm unit because it is included in the Radiation Monitoring System (RMS) control room annunciator. The detector data also are included in the RMS computer data base. An ARMS functional block diagram is shown in Figure 12.3-59. The ARMS is a part of the plant RMS and the detector data are analyzed, processed, displayed, and recorded, as described in Section 11.5. Each detector channel has three alarms: high, alert, and low (or failed). The failure alarm is adjustable and is activated if the detector, high voltage, signal, or power source fail. The high and alert setpoints are adjustable, and the local audible and visual alarms (at or near the detector) are actuated when the radiation flux exceeds the setpoint. The alarm setpoints can be changed only from the DKP or local radiation processor (LRP) and only after password/keylock entry. The LIUs have meter readouts with a five decade logarithmic scale that indicates the flux at the detector in mR/h. All channels have a five-decade range as appropriate for their detector locations, and as specified in Table 12.3-9. With the exception of the Geiger-Muller detector tubes, all electronics are solid state, and the system is designed for high reliability. All ARMS detectors are independent, and failure of one detector has no effect on any other. However, some LRPs may have more than one ARMS detector connected, and failure of the LRP will result in the loss of data from all of the detectors connected to it. The location of each area radiation detector is indicated on instrument location drawings and is listed in Table 12.3-9. The detectors can be removed from the wall and can be used as portable probes within the limit of the 25-ft detector cable to locate radiation sources. Consistent with the criteria above, the following general areas are monitored: 12.3-47 HCGS-UFSAR Revision 6 October 22, 1994
- 1. Main control room 2. Radwaste building corridors and processing areas 3. Restricted machine shop 4. Fuel storage and handling area 5. Containment hatchways and airlocks 6. Sampling stations The ARMS data processing, display, and control functions are in the CRP and DKP, which are on the plant instrument bus that is powered from a battery through an inverter. The field located detectors, LIUs, and LRPs are powered from the plant instrument bus that is powered from a battery through an inverter. No part of the ARMS is on emergency power. 12.3.4.1.5 Safety Evaluation The ARMS is not essential for safe shutdown of the plant, and it serves no active emergency function during operation. The ARMS does serve to warn plant personnel of high radiation levels in various plant areas. All ARMS detectors and LIUs are independent, and failure of one detector/LIU has no effect on any other. The ARMS is designed to operate unattended for extended periods of time, detecting and measuring ambient gamma radiation. Ambient radiation exposure rate at the detector is indicated remotely in the main control room on the DKP and also at or near the detector. These monitors cause an audible and visual alarm at the detector, and in the main control room, if the radiation levels exceed preset limits. The ARMS has no post accident monitoring function. The Drywell Atmosphere Post-Accident Radiation Monitoring System (DAPA RMS) 12.3-48 HCGS-UFSAR Revision 0 April 11, 1988 detectors are listed on Table 12.3-9 for information only and are discussed in Section 11.5.2.1.5. The ARMS is powered from battery backed 120 V ac uninterruptible instrument bus power and is not affected by a loss of offsite power. 12.3.4.1.6 Calibration and Testing Each of the ARMS monitors is calibrated by the instrument manufacturer prior to shipment using sources certified by or traceable to the National Bureau of Standards (NBS). Inplant calibration uses a standard radioactive point source traceable to NBS. The proper functioning of each ARMS monitor is verified periodically by checking instrument response to the remotely operated radioactive check source provided with each detector. The check source is located inside the detector and can be actuated from the DKP or the LRP. The trip point settings are tested by manually entering simulated data and observing the channel operation. All trips latch and must be reset manually. 12.3.4.2 Airborne Radioactive Materials Monitoring 12.3.4.2.1 Samples Taken by Hand for Laboratory Analysis Concentrations of airborne radioactive materials are determined by routine laboratory analysis of grab samples taken throughout the plant. Filter papers for particulates and charcoal cartridges for halogens are used in low volume samplers. Filter papers for particulates are used in high volume samplers. Both ventilation exhaust ducts and work areas are sampled. 12.3-49 HCGS-UFSAR Revision 12 May 3, 2002 12.3.4.2.2 Sampling Systems with Monitors Refer to Section 11.5.2.2 for additional information The ventilation systems that exhaust to the environment, north plant vent, south plant vent, and Filtration, Recirculation, and Ventilation System (FRVS) are monitored by taking a representative sample with an isokinetic probe at rates that follow the exhaust flow. The sample transport tubing is heat traced to prevent condensation. The north and south plant vent particulate filters and noble gases are monitored continuously with beta radiation detectors. The north and south plant vent halogen charcoal cartridges are monitored continuously by gamma detectors. During operation of the FRVS, only noble gases are monitored continuously with beta radiation detectors. The particulate filter and halogen charcoal cartridge are removed periodically for laboratory analysis onsite. Some of the ducts that are tributary to the normal effluent release stacks are monitored by beta radiation detectors that are inserted into the vertical wall of the duct at appropriate places. These tributary monitors will assist in locating sources of airborne radioactive materials. Sampling taps are located in the ducts next to the detectors so that grab samples can be taken. Additional mobile samplers with monitoring detectors are provided for use if needed. The above described airborne radioactive material monitoring equipment and procedures are used to meet the applicable parts of Regulatory Guides 1.21, 1.97, 8.2, 8.8, and ANSI N13.1-1969. Acceptance Criteria II.B.17 of standard review plan 12.3-12.4 provides criteria for the establishment of locations for fixed continuous area gamma radiation monitors. The specific document referenced is ANSI/ANS-HPSSC-6.8.1-1981. The locations and numbers of monitors used at HCGS are not in full compliance with 12.3-50 HCGS-UFSAR Revision 0 April 11, 1988 this standard. The locations of these monitors are in the vicinity of personnel access areas only. These locations are based on the dose assessment and operating experiences from other nuclear power plants. In addition, these locations were finalized prior to the issuance of this standard and provide an acceptable method of monitoring area radiation values.
Acceptance Criterion II.4.b.3 requires ventilation monitors to be placed upstream of the HEPA filters. The HCGS design places scintillation detectors in ducts that are tributary to the release vent in order to provide warning of increased releases within the plant. These instruments detect increases in the gross noble gas concentrations of the effluent. Hence, placement of the detectors relative to HEPA and/or charcoal filters does not significantly affect their response. Since releases of iodines and particulates will be accompanied by much larger releases of noble gases, the changes in ventilation monitor readings provide indication of a change in airborne activity concentration in one or more of the plant's areas. If an increase is detected, its source and
magnitude will be determined using portable samplers.
Normally occupied non
-radiation areas in the plant do not have potential for significant airborne concentrations of particulates and iodine during plant
operation because:
- 1. The ventilation systems are designed to prevent the spread of airborne radioactivity into normally occupied areas.
- 2. Highly radioactive piping/components are not located in normally occupied areas.
Certain activities, such as refueling, solid waste handling, or turbine teardown, may increase the possibility of encountering significant airborne activities in some normally occupied areas. Continuous local airborne
monitoring will be provided during these activities as needed.
Exposure of personnel to high concentrations of airborne activity in radiation areas will be prevented through in-plant surveys and portable particulate and iodine sampling monitors prior to personnel entrance. Continuous monitoring will be provided as required by area conditions and the nature of the entry. The locations of portable monitors (see 1 through 12 below), capable of detecting 10
DAC-hours of particulates and iodines, which are positioned within the station during normal operations to provide supplemental inplant monitoring of particulates and iodine levels.
12.3-51 HCGS-UFSAR Revision 22 May 9, 2017
- 7. Service Rad Waste Building 102 8. Service Rad Waste Building 5 (2 provided)
- 9. Turbine Building 137 (2 provided)
- 11. Turbine Building Hallway 77 Fifteen Continuous Air Monitors (CAM) are situated throughout the facility to measure, indicate and record the levels of airborne radioactivity at locations where significant airborne radioactivity is likely. Each CAM has the ability to activate a local alarm when a predetermined level is exceeded. If a CAM is broken or otherwise out of service, a Low Volume air sample can take its place
until another CAM can be placed in service.
During outages or special evolutions, these monitors may be augmented or shifted, as needed. The quantity required during normal operations and monitor type are identified in Table 12.5-1. Administrative control will prevent inadvertent entry of personnel into normally unoccupied areas (Zone III and above). The provisions discussed above ensure that personnel will not be
inadvertently exposed to significant concentrations of airborne activity.
12.3.5 References
12.3-1 J.J. Martin and P.H. Blichert-Toft, "Radioactive Atoms, Auger Electrons, and X-Ray Data," Nuclear Data Tables, Academic Press, October 1970.
12.3-2 J.J. Martin, Radioactive Atoms Supplement 1," ORNL 4923, Oak Ridge National Laboratory, August 1973.
12.3-3 W.W. Bowman and K.W. MacMurdo, "Radioactive Decays Ordered by Energy and Nuclide," Atomic Data Nuclear Data Tables, Academic Press, February 1970.
12.3-4 M.E. Meek and R.S. Gilbert, "Summary of X-Ray and Gamma-Ray Energy and Intensity Data," NEDO -12037, General Electric, January 1970.
12.3-52 HCGS-UFSAR Revision 22 May 9, 2017
13.3-5 C.M. Lederer, et al, "Table of Isotopes," 6th edition, John Wiley, New York, 1967 (1st corrected printing March 1968). 12.3-6 D.S. Duncan and A.B. Spear, "Grace 1 - An IBM 704-709 Program Design for Computing Gamma Ray Attenuation and Heating in Reactor Shields," Atomics International, NAA-SR-3719, June 1959. 12.3-7 D.S. Duncan and A.B. Spear, "Grace 2 - An IBM 709 Program for Computing Gamma Ray Attenuation and Heating in Cylindrical and Spherical Geometries," Atomics International, NAA-SR-4649, November 1959. 12.3-8 W.W. Engle, Jr, "A User's Manual for ANISN: A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," K-1693, Union Carbide Corporation, 1967. 12.3-9 E.D. Arnold and B.F. Markewitz, "SDC, A Shielding-Design Calculation for Fuel-Handling Facilities," ORNL-3041, Oak Ridge National Laboratory, March 1966. 12.3-10 R.E. Malenfant, "QAD, A Series of Point-Kernel General-Purpose Shielding Programs," LA 3573, Los Alamos Scientific Laboratory, October 1966. 12.3-11 D.A. Klopp, "NAP - Multigroup Time-Dependent Neutron Activation Predication Code," IITRI-A6088-21, January 1966. 12.3-12 E.A. Straker, et al, "MORSE - A Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code," ORNL-4585, Oak Ridge National Laboratory, September 1970. 12.3-53 HCGS-UFSAR Revision 0 April 11, 1988 12.3-13 W.A. Rhoades and F.R. Mynatt, "The DOT 3 Two-Dimensional Discrete Ordinates Transport Code," ORNL-TM-4280, Oak Ridge National Laboratory, September 1973. 12.3-14 M.J. Bell, "ORIGEN - The ORNL Generation and Depletion Code," ORNL-4628, Oak Ridge National Laboratory, May 1973. 12.3-15 R.E. Malenfant, "G-33: A General-Purpose Gamma-Ray Scattering Program," LA 5176, Los Alamos Scientific Laboratory, June 1973. 12.3-16 G.W. Grodstein, "X-Ray Attenuation Coefficients from 10 kev to 100 Mev," National Bureau of Standards, Circular 583, April 30, 1957. 12.3-17 D.K. Trubey, "A Survey of Empirical Functions Used to Fit Gamma-Ray Buildup Factors," ORNL-RSIC-10, Oak Ridge National Laboratory, February 1966. 12.3-18 Oak Ridge National Laboratory, "CASK - 40 Group Neutron and Gamma-Ray Cross Section Data," ORNL RSIC Computer Code Collection DLC-23, September 4, 1978. 12.3-19 W.E. Selph, "Neutron and Gamma Ray Albedos," ORNL-RSIC-21, Oak Ridge National Laboratory, February 1968. 12.5-20 Argonne National Laboratory, "Reactor Physics Constants," ANL-5800, July 1963. 12.3-21 J.F. Kircher and R.E. Bowman, ed, "Effects of Radiation on Materials and Components," Reinhold, New York, March 1964. 12.3-22 T. Rockwell, "Reactor Shielding Design Manual," D. Van Nostrand Co, New York, 1956. 12.3-54 HCGS-UFSAR Revision 0 April 11, 1988 12.3-23 C.R. Tipton, Jr, ed, "Materials", "Reactor Handbook," Vol I, 2nd edition, Interscience Publishers, New York, 1960. 12.3-24 H. Soodak, ed, "Physics," Reactor Handbook," Vol III, Part A, 2nd edition, Interscience Publishers, New York, 1962. 12.3-25 E.P. Blizzard and L.S. Abbot, ed, "Shielding," Reactor Handbook, Vol III, Part B, 2nd edition, Interscience Publishers, New York, 1962. 12.3-26 N.M. Schaeffer, ed, "Reactor Shielding for Nuclear Engineers," TID-25951, Atomic Energy Commission, Division of Reactor Development and Technology, 1973. 12.3-27 R.G. Jaeger, et al, "Shielding Fundamentals and Methods," Engineering Compendium on Radiation Shielding," Vol 1, Springer-Verlag, New York, 1968. 12.3-28 Nuclear Regulatory Commission, Regulatory Guide 8.8, Proposed Revision 4, March 1979. 12.3-29 General Electric, "Airborne Releases from BWRs for Environmental Impact Evaluations," NEDO-21159, March 1976. 12.3-30 General Electric, "Airborne Releases from BWRs for Environmental Impact Evaluations," NEDO-21159-2, 1977. 12.3-31 Nuclear Regulatory Commission, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Boiling Water Reactors," BWR - GALE Code, NUREG-0016, Revision 1, January 1979. 12.3-55 HCGS-UFSAR Revision 0 April 11, 1988
Figure F12.3-1 intentionally deleted. Refer to Plant Drawing N-1031 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-2 intentionally deleted. Refer to Plant Drawing N-1032 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-3 intentionally deleted. Refer to Plant Drawing N-1037 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-4 intentionally deleted. Refer to Plant Drawing N-1033 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-5 intentionally deleted. Refer to Plant Drawing N-1034 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-6 intentionally deleted. Refer to Plant Drawing N-1035 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-7 intentionally deleted. Refer to Plant Drawing N-1036 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-8 intentionally deleted. Refer to Plant Drawing N-1038 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-9 intentionally deleted. Refer to Plant Drawing N-1041 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-10 intentionally deleted. Refer to Plant Drawing N-1042 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-11 intentionally deleted. Refer to Plant Drawing N-1043 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-12 intentionally deleted. Refer to Plant Drawing N-1044 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-13 intentionally deleted. Refer to Plant Drawing N-1045 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-14 intentionally deleted. Refer to Plant Drawing N-1046 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-15 intentionally deleted. Refer to Plant Drawing N-1047 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-16 intentionally deleted. Refer to Plant Drawing N-1011 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-17 intentionally deleted. Refer to Plant Drawing N-1012 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-18 intentionally deleted. Refer to Plant Drawing N-1013 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-19 intentionally deleted. Refer to Plant Drawing N-1014 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-20 intentionally deleted. Refer to Plant Drawing N-1015 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.3-21 intentionally deleted. Refer to Plant Drawing N-1016 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014
Figure F12.3-64 intentionally deleted. Refer to Plant Drawing P-0047-1 SH 2 inDCRMS HCGS-UFSARRevision20May 9, 2014
Figure F12.3-66 intentionally deleted. Refer to Plant Drawing P-0046-1 in DCRMSHCGS-UFSARRevision20May 9, 2014
(Historical Information)
- 4. The use of remote, automatic weld inspection of the RPV nozzle welds. 5. Platforms at strategic locations to improve access and overall work
conditions.
- 6. The provision of quick opening shield doors at RPV nozzle and pipe penetrations through the biological shield.
- 7. Use of radiation shielding windows during routine surveillances of plant areas.
12.4.2 Airborne Radioactivity Dose Estimates for Exposures Within the Plant Structures
The estimated exposures to plant personnel from airborne radioactivity are based upon the source distributions and radionuclide concentrations presented in Section 12.2 and Tables 12.2-138 through 12.2-144. Because of the limited geometry afforded by the finite compartment sizes within the plant, personnel exposures due to noble gas immersion are expected to be insignificant when compared to inhalation exposures and therefore are not estimated.
To determine whether exposure contributions from airborne radioactive particulates are significant, an evaluation for each area is made of the ratio of total particulate Derived Air Concentration (DAC) fractions to total radioiodine DAC fractions, which is equivalent to the ratio of particulate DAC-hours to iodine DAC-hours. The evaluations performed in this section are based on DACs listed in A ppendix B of 10CFR20 as of 1988. For the turbine and reactor building enclosure areas, the particulate to iodine ratios are approximately 1:80 and 1:10, respectively, indicating that the particulate inhalation exposures are not significant in those areas.
In the radwaste areas of the Auxiliary Building, however, the particulate to iodine ratio is approximately 1:4. Since over 50 percent of the total particulate DAC fraction is attributable to Co-60, both the thyroid inhalation dose due to radioiodines
and the lung inhalation dose due to Co-60 are estimated for the radwaste areas, as the thyroid and the lung are the critical organs for iodines and Co-60, respectively.
12.4-10 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information) 12.5 RADIATION PROTECTION PROGRAM
12.5.1 Program Description
The Radiation Protection Program provides evaluation and documentation of site radiological conditions and ensures that every reasonable effort is made to
maintai accordance with requirements of 10CFR20, Regulatory Guides, and Technical Specifications. The program is designed to protect the public and plant personnel from unnecessary exposure to radiation and radioactive materials. The personnel responsible for the Radiation Protection Program are, in order of authority, the Hope Creek Site Vice President, the Plant Manager, the Radiation Protection Manager, Radiation Protection Superintendent, Radiation Protection
Supervisors, and Radiation Protection Technicians.
12.5.1.1 Authority and Responsibility
The Plant Manager is responsible for maintaining and implementing the Radiation Protection Program and receives direct reports from the Radiation Protection
Manager concerning the status of the program.
The Radiation Protection Manager (RPM) is responsible for managing the Radiation Protection department to meet station operational needs and
radiological safety standards.
The Radiation Protection Manager manages the Radioactive Material Control program and implements the ALARA program as described in administrative procedures.
Radiation Protection Supervisors are responsible for planning, conducting, and supervising daily radiation protection activities.
Radiation Protection Technicians implement the radiation protection program under the supervision of Radiation Protection Supervisors.
Radiation Protection personnel have the authority to halt any work activity when, in their professional judgment, worker safety is being jeopardized or
unnecessary personnel exposures are occurring.
12.5-1 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information)
In the absence of Radiation Protection supervision, the authorities of the above positions may be delegated in accordance with station radiation
protection procedures to qualified supervisors or technicians.
12.5.1.2 Experience and Qualifications
The Radiation Protection Manager is familiar with the design features of nuclear power stations and possesses both the technical competence to establish radiation protection programs and the supervisory capability to direct the work
of the professionals and technicians required to implement such programs.
The qualifications of the designated "Radiation Protection Manager" meet or exceed the requirements of Regulatory Guide 1.8, September 1975.
At least one member of the Radiation Protection Supervisor staff shall be designated as the backup "Radiation Protection Manager" in accordance with
paragraph 4.4.4(d) of ANSI/ANS 3.1-1981.
The Radiation Protection Supervisors are qualified in accordance with ANSI/ANS 3.1-1981. They shall have a minimum of four years of experience in applied radiation protection, including two years of experience in a nuclear power
plant or a nuclear facility.
The qualifications of the Radiation Protection Technicians meet or exceed the personnel requirements of ANSI/ANS 3.1-1981. Radiation Protection Technicians
are additionally trained and qualified in accordance with administrative
procedures.
12.5.2 Facilities, Equipment, and Instrumentation
Radiation protection facilities, equipment, and instrumentation were designed and acquired to meet the requirements of Regulatory Guides 8.3, 8.4, 8.8, 8.9, 8.12, 8.14, 8.15, 8.28, and 1.97 (specifically Table 2, Type E Variables for
Environs Radiation and Radioactivity).
12.5-2 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
12.5.2.1 Radiation Protection and Radiochemistry Facilities
12.5.2.1.1 Access Control
HCGS has two general area classifications for radiological control purposes:
the restricted area and the radiological control area (RCA). The restricted area is any area where access is controlled to protect all individuals from exposure to radiation or radioactive material. In general, the HCGS restricted area corresponds to the area inside the station security fence (protected area). The RCA, which is within the restricted area, features positive control over access, activities, and egress. Access is limited in accordance with operational requirements and individual training (in radiation protection). The RCA may include radiation areas, high radiation areas, very high radiation areas, contaminated areas, radioactive material storage areas, and airborne radioactivity areas. Entry to and exit from the permanent RCA is normally through two designated access control points, the Radiation Protection
Supervision may temporarily designate other entrances and exits for various plant conditions, as necessary. The access control points, shown on Plant Drawing P-0035-0 and Figure 12.5-3, are located at elevations 124 and 137 feet in the service and radwaste areas of the Auxiliary Building. Self-survey personnel monitoring equipment, such as automated whole-body contamination
monitors or Geiger-Mueller (G-M) type friskers, are located at the exit from
the RCA.
12.5.2.1.2 Radiation Protection Facilities
The radiation protection office and workrooms are located near the access control points. Portable radiation survey instrumentation, as well as air monitoring and sampling equipment, self-reading dosimeters, and miscellaneous radiation protection supplies, are stored in these rooms. Radiation protection equipment used for routine counting of smears and air samples, such as end window G-M counters, alpha and beta scintillation detectors, gas flow proportional counters, and/or gamma spectroscopy are located in the radiation
protection count room. The radiation protection office area is equipped for survey record keeping and RWP preparation. Respiratory and protective clothing
equipment are stored and issued in this area.
12.5-3 HCGS-UFSAR Revision 20 May 9, 2014 (Historical Information)
Decontamination facilities at the access control area consist of showers, sinks, and decontamination agents. Sinks and showers drain to tanks for processing through the liquid radioactive waste system. Large-area "pancake" end-window G-M friskers are located at these areas for personnel contamination
monitoring.
Cleaning of protective clothing will be provided by a vendor laundry facility.
Protective clothing, laundered offsite, is selectively monitored for contamination, sorted, and stored at the clean clothing issue areas or the
laundry storage room.
Equipment decontamination facilities are located at the 102-foot elevation of the Auxiliary Building. One room, consisting of an ultrasonic cleaner, will be used primarily for tools and small equipment. The other room is equipped to handle larger components. Equipment that cannot be completely decontaminated may be worked on in the restricted machine shop located adjacent to the
equipment decontamination facilities.
Both controlled and uncontrolled locker areas are located near the control points. The uncontrolled locker areas are used by workers not entering the RCA. Workers entering the RCA on a RWP change into clean protective clothing in the controlled locker room. Adjacent to the uncontrolled locker rooms are
toilets and washrooms, shower rooms, and drying areas.
12.5-4 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
Several changes have been made since Plant Drawing P-0035-0 was last issued.
These changes include use of some rooms for purposes other than shown on the drawing, and movement of some portable equipment to other rooms. These changes will not be shown on permanently issued drawings at this time since no structural changes have been made and it would not be cost effective to do so.
The following describes those changes approved at the time which differ from
Plant Drawing P-0035-0.
The current staffing plan projects 393 personnel during normal operation and
583 during outage periods. This includes clerical and management personnel with some facilities within the Administration Building for their use. A 10 percent female population is assumed in the normal operation workforce
exclusive of management and clerical personnel.
The total number of lockers planned for installation is 630 controlled and 570
uncontrolled. The capability for adding 20 controlled and 140 uncontrolled
lockers within these locker areas at a later date also exists.
The following is a current listing of the use of various rooms. These changes
resulted, in part, from the relocation of lockers and some portable equipment:
- 1. Room 3403 is designated for data communication and telephone equipment. Rooms 3415, 3416, 3434, 3437, and 3441 have been deleted to expand the access control point. Room 3446 has been converted to
an auxiliary personnel decontamination area.
- 2. Room 3447 will contain a self-contained respiratory cleaning and drying unit.
- 3. Room 3509 will be used as an RWP posting area for plant staff and outage personnel.
- 4. Rooms 3523 and 3524 will be used as the control point dress-out area. Room 3525 will be used as the Radiation Protection Technician work
area.
12.5-5 HCGS-UFSAR Revision 20 May 9, 2014
- 5. Room 3531 will be used for radiation protection supervisory offices due to the facilities for dosimetry being relocated to the In-Processing Center. (Historical Information)
- 6. Rooms 3535, 3536, 3537, and 3538 have been deleted to expand the access control point.
- 7. Room 3546 will be used for Radiation Work Permit (RWP) generation and records.
- 8. Rooms 3552 through 3560 (except 3555) will be used as facilities for females. Uncontrolled lockers in this area exceed 10 percent of the
staffing levels for normal and outage periods.
- 9. Room 3555 will be used as a Radiation Protection counting room. The redundancy of equipment between this room and room 3423 will ensure
that normal and emergency needs are met.
12.5.2.1.3 Radiochemistry Facilities
The radiochemistry facilities at the 124-foot elevation are part of the general chemistry facilities and consist of a hot chemistry laboratory, and a counting room. The counting room is surrounded by 18-inch concrete shield walls. An
emergency shower and eyewash is available in the radiochemistry laboratory.
12.5.2.1.4 The Security Center Facility
The security center facility contains the site access security control station guardhouse. This facility serves as the access and security control point to the site areas of both the Salem and Hope Creek units. Portal monitors, and/or large area "pancake" end-window G-M friskers, are maintained at this location
as the final monitoring check prior to leaving the restricted areas.
12.5-6 HCGS-UFSAR Revision 22 May 9, 2017
Plant Drawings A
-B102-0 and A-B103-0 show the access security control station.
12.5.2.1.5 In
-Processing Center
This facility is c ommon for both Salem and Hope Creek plants to support activities at both stations. This facility includes provisions for dosimetry issue and recordkeeping, whole body counting, and respirator fit testing.
(Historical Information) 12.5.2.2 Instruments and Equipment
Instrumentation for detecting and measuring radiation consists of counting room equipment, portable instrumentation, and air samplers. Capabilities for detecting alpha, beta, gamma, and neutron radiation are provided. Sufficient inventory is provided to accommodate use, repairs, and calibration. Instrument calibrations include electronic and radiation calibration of dose rate and count rate instruments, counting scalers, and portal monitors. It also includes the flow calibration of portable air samplers, such as AMS-3's, Hi-Vols, H-809 Lo-Vols, RAP and RAS pumps utilizing anemometers, flow meters and D/P gauges.
Electronic calibrations are performed utilizing various types of measuring and test equipment, such as current pulsers, digital volt meters, oscilloscopes, decade boxes and capacitance testers.
Radiation calibrations are performed utilizing several licensed and exempt radiation sources, depending on the type radiation the instrument detects. For gamma survey instruments, Cs-137, with an energy of 662 KeV, is the isotope chosen because it is comparable to the average gamma energy of the isotopes
present in a typical primary reactor coolant sample.
12.5-7 HCGS-UFSAR Revision 22 May 9, 2017
Three Cs-137 sources are used for calibrations. The irradiators used are the 100 mCi Shepherd Model 28-5 and a 400 Ci Shepherd Model 89 box type, which is primarily used for instruments, and a 3 Ci Shepherd Model 81-8A tower, which is used to irradiate Optically Stimulated Luminescent (OSLs) and self-reading dosimeters (SRDs).
(Historical Information)
All irradiation sources are shielded consistent with the station ALARA program.
These sources are controlled and secured under the appropriate administrative, radiation protection and physical security procedures.
Beta measuring survey instruments are calibrated and their efficiencies for beta detection determined with a depleted Uranium slab. This mylar covered slab has a standard dose rate and is commonly used for this application.
Beta counting equipment is radiation calibrated using electroplated Tc-99 planchet-sized sources.
Alpha counting scalers and alpha survey instruments are calibrated with electroplated sources, as well. These are either planchet sized or provided in detector sized jigs.
Neutron measuring instruments are calibrated with an encapsulated MRC 5.7 Ci AmBe source which is setup for irradiation on the source range in a borated polyethylene block jig.
Various low activity exempt check-sources ("button-type") are also used to verify instrument response.
All instruments are calibrated in compliance with the American National Standards Institute (ANSI) N323 - 1978, which establishes calibration methods for portable radiation protection instruments used for the detection and
measurement of ionizing radiation and radioactive surface contamination.
12.5-8 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information)
Sufficient chemical supplies, chemistry laboratory equipment, and analytical instruments are available to perform the required sample preparations and
analyses in support of radiation protection functions.
12.5.2.2.1 Chemistry Laboratories
The chemistry facility consists of two laboratories: one that handles low level or background level samples and a second that handles medium and high radiation
samples.
The laboratories are equipped with constant air flow fume hoods. The fume hoods permit preparation and processing of contaminated samples under
controlled conditions.
12.5.2.2.2 Counting Rooms
Plant system samples processed in the chemical laboratories for activity analysis and isotopic identification are transported to the chemistry counting room. Samples direct from the plant, such as air samples and smears, are transported to the radiation protection counting room. Equipment is available in both counting rooms for gross alpha, gross beta, and gross gamma activity measurements and for determination of the activity levels of specific isotopes.
Both counting rooms are temperature controlled and the voltage supply is
regulated for instrument stability.
Major instrumentation in both counting rooms includes a computer programmed multi-channel analyzer using germanium detectors, or other detectors appropriate for specific isotopic identification; gas flow proportional counters, with and/or without windows, for filter samples, smears, and planchetted water samples; and a gross beta-gamma counter. The chemistry counting room also contains a liquid scintillation counter for beta emitters
such as tritium and carbon-14.
12.5-9 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
Background and efficiency checks are performed routinely. Counter plateaus are established to determine operating voltages. Calibrations for the isotopes are based on National Bureau of Standards (NBS) related guidelines. Conventional
radionuclide reference standards will be used for calibrations.
12.5.2.2.3 Portable Survey Instruments and Equipment
Portable survey equipment is used primarily for conducting area surveys and for monitoring personnel throughout the plant. Some portable equipment is reserved for emergency use and is located in lockers at the access control point, the control room, the technical support center, and the offsite emergency operations facility.
The criteria for selection of the portable instruments include:
- 1. Ability of instrument to perform in its intended use with reliability
and accuracy
- 2. Ease of calibration and repair
- 3. Interchangeability of components
- 4. Weight and size for user acceptance
- 5. Standard readouts and controls/adjustments to simplify training of
users.
Portable instruments for routine plant use are provided to permit alpha, beta, gamma, and neutron radiation measurements, and for obtaining samples of surface and airborne contamination. Portable instruments for emergency use are provided to permit alpha, beta, and gamma radiation measurements for obtaining samples of surface and airborne contamination. PSEG Nuclear maintains and recommendations, ANSI standards, and requisite regulations. Examples of types of Health Physics equipment include: gas flow Proportional counter, electronic dosimeter, area radiation monitor (ARM), ion chamber, dose rate meter, GM dose rate meter, neutron monitoring equipment, whole body contamination monitor, GM survey count rate instrument, alpha monitoring equipment, tool and equipment monitor, air sampler, and continuous air monitor.
The frequency and methods of calibration are described in the applicable procedures.
12.5-10 HCGS-UFSAR Revision 22 May 9, 2017
12.5.2.2.4 Personnel Dosimetry Personnel monitoring will be provided by the use of Optically Stimulated
Luminescent (OSL) dosimeters, electronic dosimeters (EDs), direct reading pocket dosimeters, or calculations from area survey data and exposure times.
Personnel monitoring is provided per 10CFR parts 20 and 34. The form of personnel monitoring depends on the type of radiation and the expected radiation level. Table 12.5-2 lists the planned quantities, sensitivities, and ranges of the TLDs and self
-reading dosimeters to be used at HCGS as of 1989.
OSLs are normally used as the D osimeter of Legal Record (DLR) for individuals who require monitoring.
OSLs are used for beta, gamma, and neutron exposures and are normally processed off-site by a vendor service and evaluated on site.
Whenever neutron dose determination is required, calculations using the area neutron dose rate or area neutron to gamma ratio and dosimeter readings may be used as a backup or a tracking method prior to permanent dosimetry processing, or may be used in place of neutron dosimetry if it is more accurate. Radiatio n protection practices include the movement of the normal whole-body badge or the use of multiple badges, in addition to the whole-body badge, when exposure to specific parts of the body may be greater than the general area dose rate.
These badges will be issued at the access control point, when required by a
RWP. Electronic dosimeters are normally used to monitor gamma exposure. The results are used for specific ALARA job exposure evaluation, as well as to indicate current individual exposure status. Electronic dosimeter readings can also be used as permanent records, especially for individuals who do not require monitoring, or in case of lost or compromised DLR results. Electronic dosimeters are available at the access control point. Electronic dosimeters are response checked prior to issue, and are calibrated in accordance with
industry standards.
12.5-11 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information)
Internally deposited radioactive material is evaluated with a whole body counter. The counter is sufficiently sensitive to detect in the thyroid, lungs, or gastrointestinal tract a fraction of the annual limit on intake of the relevant gamma emitting radionuclides. The whole body counter is calibrated on an annual basis using phantoms and standard sources containing various radionuclides covering the range of energies normally expected. The detectors are used in conjunction with a multi-channel analyzer, a computer, and printer, to obtain a permanent record. Lapel air samplers are available to
aid in the assessment of airborne radiological environments.
12.5.2.2.5 Miscellaneous Instruments and Equipment
The following miscellaneous radiation protection equipment are available at one
or more locations in the plant: contamination control supplies such as glove bags, containment tents, absorbent wipes, absorbent paper, rags, step off pads, rope, plastic sheets, plastic bags, tape, contaminated area signs, and protective clothing. Appropriate supplies are assembled into kits and situated throughout the plant to aid in the control of contaminated spills. Temporary shielding, such as concrete blocks, lead bricks, lead sheets, and lead wool
blankets, is also available to reduce radiation levels.
An apparatus for quantitative fit testing of individuals involved in the
respiratory protection program is available.
Portal monitors, whole-body friskers, and/or hand-held friskers with sensitive large-area "pancake" end-window G-M probes are positioned at the RCA exit points. The purpose of these devices is to control the spread of
contamination. Other devices that prove to be of equal or greater sensitivity will be considered for use instead of those listed above. Personnel may use
friskers within the plant to monitor themselves at any time, especially when leaving a contaminated area. Portal monitors, with friskers as backup, are
used within the guardhouse as personnel leave the restricted area.
12.5-12 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
Portable ventilation systems equipped with HEPA filters or HEPA and sorbent filters, are available to minimize airborne contamination in highly
contaminated areas.
Continuous air monitors (CAMs) monitor airborne concentration at specific work locations. These CAMs record trends or sudden changes in the airborne
concentrations. While they are not intended for quantitative analysis, the fixed filter type can be used as a low volume grab air sample. The filter medium is removed and analyzed in more detail in the radiation protection
counting room.
12.5.2.2.6 Personnel Protective Equipment
Special protective equipment such as coveralls, plastic suits, shoe covers, gloves, head covers, and respirators, including approved air purifying respirators, self-contained breathing apparatus (pressure demand), and airline respirators and hoods, are stored in various plant locations and clothing change areas. This equipment is used to prevent both deposition of radioactive
material internally or on body surfaces and the spread of contamination. Most areas of the plant are kept free of contamination so that no special protective equipment is needed. Contaminated areas are identified with posted signs.
Radiation signs and radiation work permits (RWPs) are the primary means of
defining the equipment required to enter these contaminated areas.
A variety of combinations of protective equipment may be prescribed, depending on the nature and level of possible contamination. For example, cotton clothes may be adequate, but in wet areas, plastic rain suits or bubble suits may be prescribed. Respirators may be used if airborne hazards exist, or if surface contamination could cause an airborne hazard as defined in the radiation
protection procedures.
12.5-13 HCGS-UFSAR Revision 17 June 23, 2009
The use of Delta Protection Mururoa V4 F1 and V4 MTH2 respiratory protection suits has been authorized for use at Hope Creek with an assigned protection factor (APF) of 2,000 (Reference NRC to PSEG letter: "HOPE CREEK GENERATING STATION AND SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2 -REQUEST FOR AUTHORIZATION TO USE RESPIRATORY PROTECTION EQUIPMENT (TAC NOS. MD9199, MD9200, AND MD9201)", dated January 27, 2009). Approval was based on testing which demonstrated the suits met European Standard EN 1073-1 (January 1998),
"Protective Clothing Against Radioactive Contamination, Part 1: Requirements and Test Methods for Ventilated Protective Clothing Against Particulate
Radioactive Contamination." This standard is generally consistent with the pertinent acceptance criteria provided in Los Alamos National Laboratory Report
LA-10156-MS, which is used to test and authorize the use of air-supplied suits
at Department of Energy sites. The certification-testing was broadly based
covering a range of various functional areas. Both models passed all required
tests, and both provided a measured average protection level (fit factor) of 50,000. The following information on the Mururoa V4 F1 and V4 MTH2 suits is
included to comply with commitment CM.CC.2008-121:
The manufacturer's instructions for use and storage of the Delta Protection Mururoa V4F1 and V4 MTH2 suits will be adhered to and integrated into the respiratory protection program, with the exception of the requirement to have a stand-by rescue person. New lesson plans will be developed to train workers on Mururoa's features, donning, use and removal, cautions and use of mouth strip and tear off strips for routine and emergency egress. Radiation Protection personnel will be provided additional training for selection, approval, issue, equipment set-up, operation and maintenance instructions for the Mururoa suit.
The Mururoa V4F1 and V4 MTH2 suits will be discarded after a single use and will not be used in atmospheres that are immediately dangerous to life and health (IDLH). Any defects discovered with the Mururoa suit will be entered into the Corrective Action Program and reported to the manufacturer, as
necessary. Industry notifications, when required, will be made through the
Operating Experience Program.
12.5-13a HCGS-UFSAR Revision 17 June 23, 2009
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12.5-13b HCGS-UFSAR Revision 17 June 23, 2009 (Historical Information) 12.5.3 Procedures
Radiation protection procedures, as described in this section, are implemented by Hope Creek radiation protection instructions, administrative procedures, ALARA procedures, and emergency plan procedures. The procedures are written to meet the guidelines of Regulatory Guides 1.8, 1.33, 1.39, 8.2, 8.7, 8.8, 8.9, 8.10, 8.13, 8.20, 8.26, 8.27, 8.29 and Hope Creek Technical Specifications.
12.5.3.1 Radiological Surveys
Area survey procedures describe the purpose and techniques of detecting and measuring levels of radiation and contamination. Contamination may be on surfaces or airborne. Area surveys are conducted throughout the plant. Such surveys may be routine or may be related to specific jobs. An area survey may be performed before, during, and/or after various work activities. Area
surveys are performed by radiation protection personnel.
12.5.3.1.1 Radiation Detection
The preferred instrument for beta-gamma dose rate measurements is an ion chamber. G-M probes are preferred for measurement of low radiation levels or where environmental conditions such as temperature or humidity could cause
erratic responses from ion chambers.
The preferred neutron measurement instrument is a rem counter, or equivalent, that has the ability to measure neutron dose rate in rem per hour.
These radiation detection methods are supplemented by continuous area and process radiation monitoring equipment with alarm capabilities, as described in
Sections 11.5 and 12.3.4.
12.5-14 HCGS-UFSAR Revision 16 May 15, 2008 12.5.3.1.2 Surface Contamination Detection A variety of techniques are used to detect and measure radioactive contamination. Procedures prescribe the use of smears (small paper discs) and large area wipes (Maslin-type cloths) to wipe a surface to pick up removable contamination. Fixed contamination is determined by scanning a surface with
portable survey meters.
(Historical Information)
G-M probes are used for beta-gamma measurements and alpha detectors are used to
distinguish the alpha component.
12.5.3.1.3 Airborne Contamination Airborne contamination is normally determined by using air samplers to draw a known volume of air through a filter paper or charcoal cartridge. A charcoal cartridge is used with filter paper where iodine is of concern. The filter paper and charcoal cartridge are analyzed by gross beta-gamma count and/or gamma spectrometry. The gamma spectrometry identifies the particulate and radioiodine isotopic activity. Gross beta-gamma count data is used to judge the need for gamma spectrometry. High volume air samplers and low volume air samplers having nominal sample rates of 25 scfm and 2 scfm, respectively, are available. The high volume air sampler is used primarily to quickly obtain grab samples before, during, and after work activities. The low volume air sampler is used primarily to obtain the average air concentration for the work
period.
In-plant sampling for radioiodine will be performed in accordance with the PSEG Nuclear Emergency Plan and Emergency Plan Implementing Procedures.
Equipment to perform this in-plant radioiodine sampling will be kept in Emergency Lockers or in the vicinity of Emergency Lockers, in accordance with Section 9 of the Emergency Plan and Emergency Plan Inventory Procedure(s).
12.5-15 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information)
Prior to analysis, all silver zeolite cartridges analyzed inplant will be purged using bottled nitrogen gas or clean air (i.e., free of noble gases) to ensure absence of noble gases (i.e., xenon). Purging the cartridges will be
performed in a well ventilated area or under a laboratory hood.
Analyses of in-plant and onsite silver zeolite cartridges for radioiodine will normally be performed using high purity germanium detectors (HPGe). These detectors will be located in the Chemistry Laboratory (124 ft el) and the Radiation Protection Count Room (137 ft el). Analyses of offsite silver zeolite cartridges will be performed using the HP210 frisker probe with an E-140 portable count rate meter or equivalent. The HPGe detectors at the Salem Station Chemistry Laboratory may be used to analyze silver zeolite cartridges
for in-plant or onsite samples should the background radiation levels in the Chemistry Laboratory or Radiation Protection Count Room be too high to perform
analyses with the HPGe.
12.5.3.1.4 Survey Frequency and Techniques
Each area found to have a radiation dose rate such that an individual could receive 5 mrem in any one hour, is conspicuously posted as a radiation area, in accordance with 10CFR20. Every reasonable effort is made to minimize
inadvertent entries into such areas. Routine surveys of all radiation areas are taken to ensure that each area is surveyed on a regular basis. Areas subject to variations in radiation levels and occupancy times may be surveyed
on a more frequent basis. When reactor conditions are operationally stable, survey frequency in radiation areas may be reduced to spot checks at boundaries
to minimize radiation protection personnel exposures.
12.5-16 HCGS-UFSAR Revision 16 May 15, 2008
(Historical Information)
Each area found to have a radiation dose rate equal to or greater than 100 mrem/h is posted as a high radiation area and access is controlled in accordance with 10CFR20 and technical specifications. Routine surveys within such areas are not normally performed with conventional portable survey instruments. Every reasonable effort is made to use readings from the Radiation Monitoring System (RMS) area radiation monitors to identify changes of radiation levels. Measurement of maximum and general radiation levels within high radiation areas is normally performed with remote probe survey
instruments, long reach survey instruments, retrievable TLDs, or dosimeters.
When practicable, findings from these surveys are correlated to the appropriate RMS area radiation monitor readings and reactor operating conditions.
Correlation readings and/or perimeter readings are taken to ensure that each high radiation area is surveyed on a regular basis. In addition, the frequency
of radiation surveys taken at entrances to high radiation areas is dependent upon occupancy in the vicinity and variation in radiation levels. If surveys at entrances or RMS readings show significant change, additional surveys may be performed to update the surveys for the area. In order to minimize occupational exposure of surveyors, high radiation area survey frequency may be
reduced when operating conditions are stable.
Areas in and around the RCA not considered potential radiation areas are selectively surveyed to establish that every reasonable effort has been made to
keep measurable radiation at ALARA levels. Portable instrument surveys are
performed to ensure that a representative number of non-radiation areas are surveyed once per month. Areas subject to significant radiation change or
variation are surveyed on a more frequent basis, as appropriate. Any area not previously noted that is found to be a radiation area is promptly posted with a "Caution Radiation Area" sign and reported to Radiation Protection Supervision.
If the radiation field cannot be eliminated, every reasonable effort is made to minimize the dose rate and inadvertent entry. The area is placed on the
radiation area survey list.
12.5-17 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
Areas within the Hope Creek security fence not covered by normal portable instrument surveys are selectively monitored by area TLDs to document integrated exposures and are supplemented by randomly scheduled instrument
surveys.
Procedures for area surveys describe the use of instruments, effective survey techniques, and documentation of data. Procedures allow consideration for
potential as well as actual radiological hazards.
12.5.3.2 Radiation Work Permits
Where radiation dose rates, airborne concentrations, or surface contamination levels exceed station administrative control levels, a radiation work permit (RWP) is issued prior to scheduled work. This is accomplished by submittal of a RWP request form to radiation protection. Radiation protection evaluates the radiological conditions associated with the work to be performed. Based on this analysis, radiation protection may issue an extended radiation work permit if the work is of a repetitive nature and no significant short term
accumulation of person-rem is expected. Radiation protection specifies the
appropriate protective clothing, equipment, and monitoring, including dosimetry. Work area survey frequency is established by radiation protection.
All personnel performing work under a particular RWP must be familiar with permit conditions and must sign in on a register sheet. The register sheet identifies the individual, his time in and out, and his self-reading dosimeter value in and out. Radiation protection may terminate a RWP if radiological
conditions change.
Radiation protection supervision selectively reviews completed RWPs before they are filed. RWPs serve as a data source for dose comparison on repeat jobs and
can be used to determine the effectiveness of ALARA efforts.
An effective RWP program facilitates reporting requirements such as the annual
report required by Hope Creek Technical Specifications.
12.5-18 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information) 12.5.3.3 Handling and Storage of Radioactive Material
Radiation protection personnel are notified of intended releases from the RCA shipment and/or receipt of radioactive material, special nuclear material, or potential radioactive material. This is done to ensure that the radiation
protection personnel are aware of all radioactive materials onsite, so that required surveys can be performed, and to verify that correct labeling, placarding, and documentation has been performed on all potential radioactive
materials.
Calibration sources for radiation instrumentation and sources used to prepare secondary standards are stored in vault(s). These vaults are kept locked. The
locks are under the control of radiation protection management.
Small quantities of sealed or unsealed sources may be stored for convenience in
shielded cabinets, caves, or safes in secured areas. Such sources are used in the chemistry laboratories, instrument calibration facilities counting rooms, or when checking instruments' response throughout the plant.
12.5.3.4 Whole-Body Counting
Whole body counting is normally performed on personnel recently exposed to radioactivity who begin employment on site. Annual recounts may be performed.
Whole body counts may be performed when nasal or facial contamination is discovered, after a suspected internal exposure, or as determined necessary by
a Radiation Protection Supervisor.
In addition to whole-body counting, urinalysis and fecal analysis may be used for a more definite analysis of actual internally deposited radioactive
material.
12.5-19 HCGS-UFSAR Revision 16 May 15, 2008 12.5.3.5 Control of Access and Stay Time in Radiological Areas The security checkpoint at the fence line perimeter is a continuously staffed central guardhouse. Individuals assigned an OSL ensure that the proper OSL is worn upon entering the Protected Area.
(Historical Information)
Any individual without clearance to enter the restricted area must be accompanied by a person who is authorized to do so. The training, retraining, and testing requirements for unescorted access are provided in
Section 12.5.3.6. 12.5.3.5.1 Control of Radiation and High Radiation Areas Radiation, high radiation, and very high radiation areas are identified by posted radiation signs. Supplemental signs may be used to inform individuals of requirements for entry in addition to RWP requirements. Where appropriate, yellow and magenta rope or tape is used to limit access or to divert personnel to a specific control point for access. RWPs are used to describe work activities in an area, to prescribe radiation protection clothing and equipment, and to document entry and exit of each individual. Station procedures describe the purpose and application of the RWPs. Administrative guidelines for personnel exposure are established by procedure. The initial administrative control level is nominally 2000 mrem per year, which is less than the exposure dose limits in 10CFR20. Deviation from these guidelines must be requested and approved. Procedures provide the steps for approval of dose extensions. Additionally, entry to high radiation areas and very high radiation areas is normally controlled by locked doors or gates. Keys for these high radiation doors are under administrative control of the radiation protection supervision and the operations superintendent. Each key used and the individual using the key are recorded. Certain work activities exposed to high dose rates may be monitored continuously to prevent personnel from inadvertently exceeding the recommended stay time determined at the start of such work. Personnel are advised to observe the reading on their electronic dosimeter frequently.
12.5-20 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information) 12.5.3.5.2 Contamination Control
Radioactive contamination can exist in radiologically controlled areas. Access to contaminated areas is confined to specific locations. Floor coverings called step off pads define these locations. Personnel wishing to enter contaminated areas must review the RWP to determine the radiation protection clothing required. The RWP or related survey data sheets contain stay time determining information based on actual or potential airborne direct radiation, and/or contamination levels. Derived Air Concentration (DAC) hour calculations
are maintained for each individual whose internal exposure meets or exceeds administrative assessment criteria. These calculations include respiratory
protection factors when applicable. Procedures provide instructions for selection and application of protective clothing and respirators under various specific conditions. Procedures are also established for inspection, cleaning, and maintenance of such protective equipment. As personnel leave the work area, they remove the protective clothing and respirators before stepping onto the step-off pad. Personnel must frisk themselves to ensure that no contamination has been transferred to their bodies or clothing. If the frisker alarms, radiation protection is required to be notified. Radiation protection will take appropriate actions to minimize further spread of contamination, and
direct appropriate decontamination of effected areas and personnel.
When personnel contamination is noted, a radiation protection investigation appropriate to the incident will be performed. A contamination incident found to have caused a suspected intake of radioactive material will be promptly reported to appropriate supervision. When applicable, recommended methods to prevent recurrence will be forwarded to the Plant Manager for concurrence and
implementation by his directive.
Potentially contaminated tools, trash, and equipment to be transported from a contaminated area must be surveyed or bagged at the step off pad. Contaminated material must be placed in an externally non-contaminated container and appropriately labeled. The material may be placed in storage, taken to another contaminated area for reuse, or designated for appropriate disposal or
decontamination. Clean material may be released for use within the station.
Refer to Section 12.5.3.3 for unconditional release if the material is to be
removed from the restricted area.
12.5-21 HCGS-UFSAR Revision 16 May 15, 2008 (Historical Information)
The presence of radioactive contamination, whether surface or airborne, inhibits mobility of personnel within the plant. Protective clothing that must be worn creates inconveniences and introduces other factors that affect performance. For these reasons, plus the obvious potential external and internal radiological hazards, decontamination is initiated judiciously to confine the contamination levels thus minimizing protective requirements.
Special coatings that aid in decontamination are applied to walls and floors.
The ventilation flow pattern is from clean areas to contaminated areas. Process equipment is isolated in various cavities or cells. Components can be vented in a controlled manner, usually through filters to the plant vent, where flow rate and radioactivity are monitored. Highly contaminated equipment drains are piped to sumps to avoid the use of floor drains and attendant spillage of
fluids onto the floor.
12.5.3.6 Radiation Protection Training Programs
Radiation protection training programs ensure that personnel who have unescorted access to the restricted areas possess an adequate understanding of
radiation protection to maintain occupational radiation exposures as low as
reasonably achievable. Special training or retraining is administered upon recommendation of the Radiation Protection Manager. The appropriate personnel
schedule the necessary training and maintain training records.
12.5.3.6.1 Employee Training
Each individual, whether employee or contractor, must receive general employee training (GET) to be eligible for unescorted access to the Restricted Area. GET covers general site personnel response to emergency situations, basic radiation protection, quality assurance program requirements, security, and a general safety orientation. Those desiring unescorted access to the RCA must also receive radiation worker training (RWT). RWT covers detailed radiation protection principles and provides radiation worker instruction in exposure controls for direct radiation, airborne and contamination sources, the purposes served by different types of protective clothing, and how to properly don, remove, and dispose of such clothing. The type of access badges to be issued are based on this documentation. Retraining is administered annually and
includes lessons learned items.
12.5-22 HCGS-UFSAR Revision 16 May 15, 2008 Individuals who have not received GET (i.e., visitors) shall be escorted by GET qualified individuals and shall receive pertinent site instructions with each entry. Training is not required for members of the public entering the Restricted Area.
(Historical Information)
Individuals who have not received RWT for unescorted access to the RCA shall be escorted by individuals who are currently so qualified and shall receive radiation protection training commensurate with the purpose for entry, prior to such entry, as determined by radiation protection supervision.
All individuals working in the Restricted Area are given instruction concerning prenatal radiation exposure, as defined in Regulatory Guide 8.13. Occupationally exposed visitors who enter the Restricted Area also receive
t hese instructions commensurate with their purpose for entry.
12.5.3.6.3 Respiratory Equipment Training and Fit Test
Certain individuals may be required to wear specific respiratory equipment in the performance of their responsibilities. A separate training class will be conducted for them in the purpose, use, and limitations of specific respiratory
protective equipment used at the site.
To be eligible for duty which requires the use of respiratory equipment, an individual must pass the respiratory fit test, must have received the respiratory equipment training, and must be medically certified as being capable of working safely while wearing respiratory equipment. A quantitative fit test will be used to prove a satisfactory respirator fit for different types of respirators. Procedures will be established that describe the technique and define acceptance criteria.
12.5.3.6.3 Radiation Protection Personnel Training
A radiation protection training program for radiation protection technicians is provided which meets the requirements of ANSI 3.1-1981. This program instructs new radiation protection technicians in operational and analytical radiation protection procedures and theories and fundamentals of radiation safety and familiarizes them with plant layout and systems. During successful completion of the modular training program for technicians, individuals receive further comprehensive instruction for the specific requirements of their positions, including training on Emergency Plan duties.
12.5-23 HCGS-UFSAR Revision 22 May 9, 2017
(Historical Information) 12.5.3.7 Radiation Protection Records Radiation protection records, which are generated from procedural requirements of Sections 12.5.3.1 through 12.5.3.6, are maintained and retained to meet regulatory and technical specification requirements.
12.5.3.8 ALARA Program
Basic ALARA philosophies, policies and responsibilities are discussed in
Section 12.1.
The PSEG Nuclear LLC ALARA program affects all elements of the radiation protection program. In addition to the ALARA program guidance in PSEG Nuclear LLC administrative procedures, many specific ALARA topics are covered in radiation protection department procedures such as the use of portable
shielding, completion of ALARA reviews, and exposure reduction methods.
The program also specifies responsibilities, requirements, and documentation for such issues as:
- 1. Pre-job planning
- 2. Job performance evaluation
- 3. Post job review
- 4. Process review
- 5. ALARA review of procedures
- 6. Station ALARA committee.
12.5-24 HCGS-UFSAR Revision 16 May 15, 2008
TABLE 12.5-1
This Table has been deleted
1 of 1 HCGS-UFSAR Revision 22 May 9, 2017
TABLE 12.5-2
This Table has been deleted
1 of 1 HCGS-UFSAR Revision 22 May 9, 2017
Figure F12.5-2 intentionally deleted. Refer to Plant Drawing P-0035-0 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014
Figure F12.5-4 intentionally deleted. Refer to Plant Drawing A-B102-0 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014 Figure F12.5-5 intentionally deleted. Refer to Plant Drawing A-B103-0 in DCRMS HCGS-UFSAR Revision 20 May 9, 2014