Text

Redacted Braidwood Station, Units 1 & 2 and Byron Station, Unit 1 & 2, Revision 17 to Updated Final Safety Analysis Report, Chapter 12, Radiation Protection
	ML19170A387
Person / Time
Site:	Byron, Braidwood
Issue date:	06/19/2019
From:	Mahesh Chawla; Plant Licensing Branch III
To:	; Exelon Generation Co
	Chawla M 415-1447
Shared Package
ML18355A456	List: ML19168A119; ML19168A216; ML19169A023; ML19169A113; ML19170A007; ML19170A012; ML19170A311; ML19170A336; ML19170A347; ML19170A384; ML19170A387; RS-18-144, Amendment 28 to Fire Protection Report, Chapter 4, Fire Protection Program;
References
	Download: ML19170A387 (315)
	v • d • e

B/B-UFSAR CHAPTER 12.0 - RADIATION PROTECTION TABLE OF CONTENTS PAGE 12.0 RADIATION PROTECTION 12.1-1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE 12.1-1 12.1.1 Policy Considerations 12.1-1 12.1.1.1 Organization Structure 12.1-1 12.1.1.2 Personnel Activities and Responsibilities 12.1-1 12.1.1.3 Administration Concerns 12.1-2 12.1.2 Design Considerations 12.1-3 12.1.2.1 Radiation Protection Design Goals 12.1-4 12.1.2.2 Facility Design Considerations 12.1-4 12.1.2.2.1 Station Layout (Shielding) 12.1-5 12.1.2.2.2 Ventilation 12.1-6 12.1.2.2.3 Health Physics 12.1-6 12.1.2.2.4 Access Control 12.1-6 12.1.2.2.5 Control of Radioactive Fluids and Effluents 12.1-7 12.1.2.2.6 Safety Objectives 12.1-7 12.1.2.3 Improvements in Facility Design Due to Past Experience and Operation 12.1-7 12.1.2.4 Equipment Design Considerations 12.1-9 12.1.2.5 Equipment Selection 12.1-9 12.1.2.6 Overall Impact of Design Considerations 12.1-10 12.1.2.7 Radiation Protection Design Review 12.1-11 12.1.3 Operational Considerations 12.1-13 12.1.3.1 Operational Objectives 12.1-13 12.1.3.2 Implementation of Procedures and Techniques 12.1-14 12.1.3.3 Implementation of Exposure Tracking and 12.1-14 Exposure Reduction Program 12.2 RADIATION SOURCES 12.2-1 12.2.1 Contained Sources 12.2-1 12.2.1.1 NSSS Sources 12.2-1 12.2.1.2 Balance of Plant Shielding Design-Basis Sources 12.2-5 12.2.1.2.1 Blowdown Sources 12.2-5 12.2.1.2.2 Radwaste Processing Sources 12.2-5 12.2.1.2.3 Spent Resin Storage Tank (Byron) 12.2-6 12.0-i REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 12.2.1.2.4 Radwaste Solidification System (Byron) 12.2-6 12.2.1.2.3 Spent Resin Storage Tank (Braidwood) 12.2-6a 12.2.1.2.4 Radwaste Solidification System (Braidwood) 12.2-6a 12.2.1.2.5 Volume Reduction System 12.2-7 12.2.1.3 Sources for HVAC Charcoal Filters 12.2-7 12.2.1.4 Old Steam Generator Storage Facility 12.2-8 12.2.2 Airborne Radioactive Material Sources 12.2-8 12.2.2.1 Production of Radioactive Airborne Material 12.2-8 12.2.2.2 Sources in Areas Normally Accessible to Operating Personnel 12.2-9 12.2.2.3 Calculated Concentrations During Operation 12.2-10 12.2.2.4 Models and Parameters Used in Calculations of Airborne Radioactivity Concentration 12.2-10 12.2.2.5 Stack Effluents 12.2-10 12.2.3 Changes to Source Data Since PSAR 12.2-10 12.2.4 Impact of Uprate on Radiation Source Terms 12.2.5 References 12.2-11 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 Description of Facility Design Considerations 12.3-1 12.3.1.1 Equipment Selection, Layout, and Segregation 12.3-1 12.3.1.2 Cubicle Access 12.3-2 12.3.1.3 Draining and Flushing Capability of Equipment 12.3-3 12.3.1.4 Floor and Sink Drains 12.3-5 12.3.1.4.1 Design of Drain System 12.3-6 12.3.1.5 Venting of Equipment 12.3-7 12.3.1.5.1 Sumps Requiring Venting 12.3-8 12.3.1.6 Routing and Shielding of Lines and Ventilation Ducts 12.3-8 12.3.1.6.1 Routing and Shielding of Lines 12.3-8 12.3.1.6.2 Routing and Shielding of Ventilation Ducts 12.3-10 12.3.1.7 Waste Filters and Demineralizers 12.3-10 12.3.1.8 Valves and Instruments 12.3-11 12.3.1.8.1 Valves 12.3-11 13.3.1.8.2 Instruments 12.3-12 12.3.1.9 Contamination Control and Decontamination 12.3-13 12.3.1.9.1 Equipment Decontamination Facilities 12.3-15 12.3.1.9.2 Personnel Decontamination Facilities 12.3-16 12.3.1.9.3 Station Decontamination 12.3-16 12.3.1.10 Traffic Patterns and Access Control Points 12.3-16 12.3.1.11 Radiation Zones 12.3-16 12.0-ii REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 12.3.1.12 Laboratory Complex 12.3-19 12.3.1.12.1 High Level Laboratory 12.3-19 12.3.1.12.2 Low Level Laboratory 12.3-20 12.3.1.12.3 Counting Room 12.3-20 12.3.1.12.4 Chemistry Storage 12.3-20 12.3.1.12.5 Mask Cleaning Room (Byron) 12.3-21a 12.3.1.12.6 Personnel Decontamination Room (Byron) 12.3-21a 12.3.1.12.7 Office Space (Byron) 12.3-21a 12.3.1.13 Laundry Facility (Byron) 12.3-21a 12.3.1.12.5 Instrument Storage Room (Braidwood) 12.3-21b 12.3.1.12.6 Personnel Decontamination Room (Braidwood) 12.3-21b 12.3.1.12.7 Office Space (Braidwood) 12.3-21b 12.3.1.13 Laundry Facility (Braidwood) 12.3-21b 12.3.1.14 Survey Instrument Calibration Room 12.3-22 12.3.1.15 Locker Room Facilities 12.3-22 12.3.1.16 Design Features to Assist Decommissioning 12.3-22 12.3.1.17 Old Steam Generator Storage Facility 12.3-23 12.3.2 Shielding 12.3-23 12.3.2.1 General Shielding Design Criteria 12.3-23 12.3.2.1.1 Regulatory Requirements 12.3-23a 12.3.2.1.2 Shielding Requirements 12.3-23a 12.3.2.1.3 Design Requirements 12.3-26 12.3.2.1.4 General Description and Design Parameters 12.3-26 12.3.2.1.5 Shielding Materials and Construction Methods 12.3-27 12.3.2.1.6 Removable Shield Walls, Portable Shielding, and Compensatory Shielding 12.3-27 12.3.2.1.6.1 Stacked (Unmortared) Block 12.3-28 12.3.2.1.6.2 Removable Shield Hatches and Plugs 12.3-28 12.3.2.1.6.3 Shield Doors 12.3-29 12.3.2.1.7 Inspection (Inservice) and Maintenance Requirements 12.3-29 12.3.2.1.8 Shield Thicknesses 12.3-29 12.3.2.1.9 Calculational Methods 12.3-30 12.3.2.2 Specific Shielding Design Criteria 12.3-31 12.3.2.3 Shield Wall Penetrations and Streaming Ratios 12.3-33 12.3.3 Ventilation Requirements 12.3-35 12.3.3.1 Station Ventilation 12.3-35 12.3.3.2 Design Criteria 12.3-36 12.3.3.3 Cubicles Requiring Charcoal Air Filtration (Byron) 12.3-38a 12.3.3.3 Cubicles Requiring Charcoal Air Filtration (Braidwood) 12.3-40a 12.3.3.4 Ventilation Design Features 12.3-40c 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation 12.3-40c 12.3.4.1 Area Radiation Monitoring Instrumentation 12.3-40c 12.3.4.2 Continuous Airborne Monitoring Instrumentation 12.3-42 12.3.5 References 12.3-45 12.0-iii REVISION 10 - DECEMBER 2004

B/B-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 12.3A EXAMPLES OF THE APPLICATION OF RADIATION PROTECTION DESIGN FEATURES TO SPECIFIC COMPONENTS 12.3A-1 12.4 DOSE ASSESSMENT 12.4-1 12.4.1 Estimated Occupancy of Plant Radiation Zones 12.4-1 12.4.2 Estimates of Inhalation Doses 12.4-1 12.4.3 Objectives and Criteria for Design Dose Rates 12.4-1 12.4.4 Estimated Annual Occupational Exposures 12.4-1 12.4.5 Estimated Annual Dose at the Exclusion Area Boundary 12.4-2 12.4.6 Deleted 12.4.7 Dose Reduction Program 12.4-2 12.4.8 Radiological Environmental Monitoring Program 12.4-3 12.4.9 References 12.4-3a 12.5 HEALTH PHYSICS PROGRAM 12.5-1 12.5.1 Organization 12.5-1 12.5.2 Equipment, Instrumentation, Facilities 12.5-1 12.5.3 Procedures 12.5-1 12.5.3.1 Administrative Program 12.5-2 12.5.3.2 Personnel External Exposure Program 12.5-2 12.5.3.3 Personnel Internal Exposure Program 12.5-3 12.5.3.4 Contamination Control Program 12.5-5 12.5.3.5 Training Program 12.5-5 12.0-iv REVISION 5 - DECEMBER 1994

B/B-UFSAR CHAPTER 12.0 - RADIATION PROTECTION LIST OF TABLES NUMBER TITLE PAGE 12.1-1 NSSS Radiation Protection Personnel 12.1-16 12.1-2 Radiation Protection Personnel Participating in the AE's Submittal Radiation Protection Design Review 12.1-17 12.2-1 Reactor Coolant Nitrogen-16 Activity 12.2-12 12.2-2 Reactor Coolant Sources for Shielding Design (Original Design Basis) 12.2-13 12.2-3 Deposited Corrosion Product Activity on Steam Generator Primary Side Surfaces

(µCi/cm2) 12.2-14 12.2-4 Pressurizer Liquid Phase Activity 12.2-15 12.2-5 Pressurizer Steam Phase Activity 12.2-16 12.2-6 Pressurizer Deposited Activity 12.2-17 12.2-7 Letdown Coolant Activity 12.2-18 12.2-8 Volume Control Tank 12.2-19 12.2-9 Recycle Holdup Tank Vapor Phase Sources (8087 ft3) 12.2-21 12.2-10 Recycle Holdup Tank Liquid Phase Sources (6886 ft3) 12.2-22 12.2-11 Recycle Evaporator Vent Condenser Section (7900 cm3) 12.2-23 12.2-12 Residual Heat Removal Loop Residual Heat Removal Loop Sources 12.2-24 12.2-13 Mixed Bed Demineralizer (30 ft3) 12.2-25 12.2-14 Cation Bed Demineralizer (20 ft3) 12.2-26 12.2-15 Thermal Regeneration Demineralizer (70 ft3) 12.2-27 12.2-16 Recycle Evaporator Feed Demineralizer (30 ft3) 12.2-28 12.2-17 Recycle Evaporator Condensate Demineralizer (20 ft3) 12.2-29 12.2-18 Spent Fuel Pit Demineralizer (30 ft3) 12.2-30 12.2-19 Reactor Coolant Filter 12.2-31 12.2-20 Seal Water Return Filter, Recycle Evaporator Feed Filter, Spent Fuel Pit Filter, and Spent Fuel Pit Skimmer Filter 12.2-32 12.2-21 Recycle Evaporator Concentrate Filter 12.2-33 12.2-22 Recycle Evaporator Condensate Filter 12.2-34 12.2-23 Core Shutdown Sources - (MeV/cm3-sec) 12.2-35 12.2-24 Irradiated Ag-In-Cd Control Rod Sources (Ci/cm/rod) 12.2-36 12.2-24a Hafnium Control Rod Source Strengths 12.2-37 12.2-25 Refueling Water Activity Concentrations Resulting in 2.5 mrem/hr At the Surface 12.2-38 12.0-v REVISION 9 - DECEMEBER 2002

B/B-UFSAR LIST OF TABLES (Cont'd)

NUMBER TITLE PAGE 12.2-26 Incore Instruments - Fission Chamber Sources 12.2-39 12.2-27 Drive Wire Sources 12.2-40 12.2-28 Single Waste Gas Decay Tank Activities 12.2-41 12.2-29 Spent Fuel Pit Water Activity for a Fuel Handling Accident 12.2-42 12.2-30 Shielding Design-Basis Influent Radioactivity Concentration in Liquid Waste Processing Streams 12.2-43 12.2-31 Shielding Design-Basis Influent R Radioactivity Concentrations in Liquid Waste Processing Streams 12.2-44 12.2-32 Source Bases for Drain Tanks 12.2-46 12.2-33 Laundry Drain Sources Used in Shielding Source Calculation 12.2-47 12.2-34 Decontamination Factors Used in Shielding Source Calculation of Liquid Radwaste Processing System and Blowdown System Components 12.2-48 12.2-35 Shielding Design-Basis Radionuclide Content in Liquid Radwaste Processing System Components 12.2-49 12.2-36 Shielding Design-Basis Radionuclide Content in Liquid Radwaste Processing System Components (Curies) 12.2-51 12.2-37 Shielding Design-Basis Radionuclide Content in Radwaste Filters (in Curies) 12.2-53 12.2-38 Shielding Design-Basis Radionuclide Content in Radwaste Filters (in Curies) 12.2-54 12.2-39 Shielding Design-Basis Radionuclide Content in Liquid Radwaste Processing System Components (Curies) 12.2-55 12.2-40 Shielding Design-Basis Radionuclide Content in Liquid Radwaste Processing System Components (Curies) 12.2-57 12.2-41 Shielding Design-Basis Radionuclide Content in Liquid Radwaste Processing System Components (Curies) 12.2-58 12.2-42 Assumed Demineralizer Resin Inventory in Spent Resin Tank for Shielding Sources Calculation (Byron) 12.2-59 12.2.42 Assumed Demineralizer Resin Inventory in High Activity Spent Resin Tank for Shielding Source Calculation (Braidwood) 12.2-59a 12.2-43 Spent Resin Tank Shielding Design-Basis Radionuclide Content (in Curies) (Byron) 12.2-60 12.2-43 High Activity Spent Resin Tank Shielding Design-Basis Radionuclide Content (in Curies) (Braidwood) 12.2-60a 12.2-44 Composition of a Single 55-Gallon Radwaste Drum for Shielding Analysis of Drum Storage Areas 12.2-61 12.0-vi REVISION 1 - DECEMBER 1989

B/B-UFSAR LIST OF TABLES (Cont'd)

NUMBER TITLE PAGE 12.2-45 Design-Basis Shielding Sources for Main Auxiliary Building Charcoal Air Filter and Off-Gas Vent Filter 12.2-62 12.2-46 Calculated Airborne Activities for Design-Basis Leak Rate in Auxiliary Building 12.2-63 12.2-47 Calculated Airborne Activities for Design-Basis Leak Rate in Containment Building 12.2-68 12.2-48 Calculated Airborne Activities for Design-Basis Leakrate in Radwaste Building 12.2-69 12.2-49 Deleted 12.2-70 12.2-50 Deleted 12.2-71 12.2-51 Deleted 12.2-72 12.2-52 Deleted 12.2-73 12.2-53 Assumed Demineralizer Resin Inventory in Low Activity Spent Resin Tank for Shielding Source Calculation (Braidwood) 12.2-74 12.2-54 Low Activity Spent Resin Tank Shielding Design-Basis Radionuclide Content (Braidwood) 12.2-75 12.2-55 Old Steam Generator Storage Facility Surveyed Dose Rates 12.2-76 12.3-1 Classification of Radiation Zones for Shield Design and Radiological Access Control 12.3-47 12.3-2 Specific Shielding Design Criteria 12.3-49 12.3-3 Area Radiation Monitors 12.3-75 12.3-4 Parameters Used in the Calculation of the Primary Shield Thickness 12.3-84 12.3-5 Core Fission Source for Primary Shield Calculation 12.3-86 12.3-6 Shielding Design-Basis Geometry for Shielding Thickness Calculations 12.3-87 12.3-7 Estimated Occupational Radiation Exposure During Decommissioning 12.3-94 12.3-8 Dominant Radioactive Isotopes for Prompt Dismantling and Delayed Dismantling 12.3-95 12.3-9 Sensitivity of Continuous Airborne Monitoring System 12.3-96 12.4-1 Personnel Exposure Data for Various Operating PWRs (1,15) 12.4-5 12.4-2 Personnel Exposure Data for Multiple-Unit Operating PWRs (1,15) 12.4-9 12.4-3 Reported Personnel Exposure by Work Function for Several Operating PWRs (5-14) 12.4-11 12.0-vii REVISION 9 - DECEMBER 2002

B/B-UFSAR LIST OF TABLES (Cont'd)

NUMBER TITLE PAGE 12.4-4 Annual Thyroid Doses Resulting from Calculated Design-Basis Airborne Concentrations in rems/yr 12.4-14 12.4-5 Estimated Fifth Year Radiation Dose for B/B Compared with 1976 and 1977 Operating Data 12.4-15 12.5-1 Storage Location of Equipment 12.5-6 12.5-2 Health Physics Equipment 12.5-7 12.5-3 Health Physics and Radiochemical Facilities 12.5-9 12.0-viii REVISION 7 - DECEMBER 1998

B/B-UFSAR CHAPTER 12.0 - RADIATION PROTECTION LIST OF FIGURES NUMBER TITLE 12.2-1 Deleted 12.3-1 Sketch of a Simple Labyrinth Entrance 12.3-2 Sketch of a Double Labyrinth Entrance 12.3-3 Typical Walk-In Valve Aisle 12.3-4 Sketch of Radiation detector Probe Access Hole in Shield Hatch for Filter Demineralizer 12.3-5 through 12.3-26a Deleted 12.0-ix REVISION 10 - DECEMBER 2004

B/B-UFSAR LIST OF FIGURES (Cont'd)

NUMBER TITLE 12.3-27 Radiation Zone Map for Normal Operation Roof Plans El.

477 ft 0 in. and El. 485 ft 0 in. Columns 18 through 30 12.3-28 Radiation Zone Map for Normal Operation Roof Plan El.

477 ft 0 in. and El. 485 ft 0 in. Columns 6 through 18 12.3-29 Radiation Zone Map for Normal Operation Auxiliary Building El. 451 ft 0 in.

12.3-30 Radiation Zone Map for Normal Operation Auxiliary Building El. 439 ft 0 in.

12.3-31 Radiation Zone Map for Normal Operation Auxiliary Building El. 426 ft 0 in.

12.3-32 Radiation Zone Map for Normal Operation Auxiliary Building El. 401 ft 0 in.

12.3-33 Radiation Zone Map for Normal Operation Auxiliary Building El. 383 ft 0 in.

12.3-34 Radiation Zone Map for Normal Operation Auxiliary Building El. 364 ft 0 in.

12.3-35 Radiation Zone Map for Normal Operation Auxiliary Building El. 346 ft 0 in.

12.3-36 Radiation Zone Map for Normal Operation Miscellaneous Plans 12.3-37 Radiation Zone Map for Normal Operation Pipe Tunnels 12.3-38 Radiation Zone Map for Normal Operation Areas Between Auxiliary Building and Containment Building 12.3-39 Radiation Zone Map for Normal Operation Fuel Handling Building El. 426 ft 0 in.

12.3-40 Radiation Zone Map for Normal Operation Fuel Handling Building El. 401 ft 0 in.

12.3-41 Radiation Zone Map for Normal Operation Containment Building El. 377 ft 0 in.

12.3-42 Radiation Zone Map for Normal Operation Containment Building El. 390 ft 0 in.

12.3-43 Radiation Zone Map for Normal Operation Containment Building El. 401 ft 0 in.

12.3-44 Radiation Zone Map for Normal Operation Containment Building El. 426 ft 0 in.

12.3-45 Radiation Zone Map for Normal Operation Radwaste/

Service Building El. 397 ft 0 in.

12.0-x REVISION 9 - DECEMBER 2002

B/B-UFSAR LIST OF FIGURES (Cont'd)

NUMBER TITLE 12.3-46 Radiation Zone Map for Normal Operation Radwaste/

Service Building El. 433 ft 0 in.

12.3-47 Radiation Zone Map for Normal Operation Condensate Polishing/Technical Support Center 12.3-48 Radiation Zone Map for Normal Operation Auxiliary Building Elevations 459 ft 2 in., 463 ft 5 in., and 475 ft 6 in.

12.3-49 Radiation Zone Map for Shutdown Roof Plan El. 477 ft 0 in. and El. 485 ft 0 in. Columns 18 through 30 12.3-50 Radiation Zone Map for Shutdown Roof Plan El. 477 ft 0 in. and El. 485 ft 0 in. Columns 6 through 18 12.3-51 Radiation Zone Map for Shutdown Auxiliary Building El.

451 ft 0 in.

12.3-52 Radiation Zone Map for Shutdown Auxiliary Building El.

439 ft 0 in.

12.3-53 Radiation Zone Map for Shutdown Auxiliary Building El.

426 ft 0 in.

12.3-54 Radiation Zone Map for Shutdown Auxiliary Building El.

401 ft 0 in.

12.3-55 Radiation Zone Map for Shutdown Auxiliary Building El.

383 ft 0 in.

12.3-56 Radiation Zone Map for Shutdown Auxiliary Building El.

364 ft 0 in.

12.3-57 Radiation Zone Map for Shutdown Auxiliary Building El.

346 ft 0 in.

12.3-58 Radiation Zone Map for Shutdown Miscellaneous Plans 12.3-59 Radiation Zone Map for Shutdown Pipe Tunnels 12.3-60 Radiation Zone Map for Shutdown Areas Between Auxiliary Building and Containment Building 12.3-61 Radiation Zone Map for Shutdown Fuel Handling Building El. 426 ft 0 in.

12.3-62 Radiation Zone Map for Shutdown Fuel Handling Building El. 401 ft 0 in.

12.3-63 Radiation Zone Map for Shutdown Containment Building El. 377 ft 0 in.

12.3-64 Radiation Zone Map for Shutdown Containment Building El. 390 ft 0 in.

12.3-65 Radiation Zone Map for Shutdown Containment Building El. 401 ft 0 in.

12.3-66 Radiation Zone Map for Shutdown Containment Building El. 426 ft 0 in.

12.3-67 Radiation Zone Map for Shutdown Radwaste/Service Building El. 397 ft 0 in.

12.3-68 Radiation Zone Map for Shutdown Radwaste/Service Building El. 433 ft 0 in.

12.3-69 Radiation Zone Map for Shutdown Condensate Polishing/

Technical Support Center 12.0-xi

B/B-UFSAR LIST OF FIGURES (Cont'd)

NUMBER TITLE 12.3-70 Radiation Zone Map for Shutdown Auxiliary Building Elevations 459 ft 2 in., 463 ft 5 in., and 475 ft 6 in.

12.3-71 Radiation Zone Map for the Old Steam Generator Storage Facility 12.3A-1 Filter/Demineralizer Equipment, Sampling Station and Panel 12.3A-2 Hydrogen Recombiner 12.3A-3 Evaporator Equipment 12.3A-4 Removable Block Wall Plan 12.3A-5 Removable Block Wall Sections DRAWINGS CITED IN THIS CHAPTER*

The listed drawings are included as General References only; i.e., refer to the drawings to obtain additional detail or to obtain background information. These drawings are not part of the UFSAR. They are controlled by the Controlled Documents Program.

DRAWING* SUBJECT M-24-1 to -23 General Arrangements, Radiation Shielding Units 1 &

2 M-48A Composite Diagram of Liquid Radwaste Treatment Processing Units 1 & 2 12.0-xii REVISION 10 - DECEMBER 2004

B/B-UFSAR CHAPTER 12.0 - RADIATION PROTECTION 12.0 RADIATION PROTECTION The design-basis shielding sources were determined using the conservative source model in Subsection 11.1.2.1.

12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLY ACHIEVABLE 12.1.1 Policy Considerations It is the policy of Exelon Generation Company to maintain occupational radiation exposure as low as is reasonably achievable (ALARA), consistent with plant construction, maintenance, and operational requirements, and within the applicable regulations. Regulatory Guide 8.8, Sections C.1, C.3, and C.4 is used as a basis for developing the ALARA and radiation protection programs with the following exceptions: C1B page 8.8 qualifications for radiation protection manager (RPM) job

- the stations do not commit to requiring the RPM to take any type of certification exam.

Exelon Generation Company ALARA policy applies to total person-rems accumulated by personnel, as well as to individual exposures. Exelon Generation Company management provides the environment for this policy to function in a proper manner.

Management's commitment to this policy is reflected in the design of the plant, the careful preparation of plant operating and maintenance procedures, the provision for review of these procedures and for review of equipment design to incorporate the results of operating experience, and most importantly, the establishment of an ongoing training program. Training is provided for all personnel (Subsection 13.2.1), so that each individual is capable of carrying out his responsibility for maintaining his own exposure ALARA consistent with discharging his duties and also that of others. The development of the proper attitudes and awareness of the potential problems in the area of health physics is accomplished by proper training of all plant personnel. The organizational structure related to assuring that occupational radiation exposure be maintained ALARA is described in Subsection 12.1.1.1.

12.1.1.1 Organization Structure The operating organization structure of the Byron/Braidwood Stations (B/B) is described in Chapter 13.0. Reporting to the Radiation Protection Department Head are health physicists, supervisors, and technicians.

The Radiation Protection Department Head is responsible for the overall radiation protection and ALARA programs and reports to 12.1-1 REVISION 17 - DECEMBER 2018

B/B-UFSAR the station manager. Periodic meetings are scheduled between the Radiation Protection Department Head and the station manager to discuss radiation protection concerns. Also, Radiation Protection Department personnel periodically meet with the ALARA committee to discuss ALARA concerns. Several station departments (e.g., operations, maintenance, station management, etc.)

participate in these meetings.

12.1.1.2 Personnel Activities and Responsibilities The station Radiation Protection Manager is responsible for the health physics program and for handling and monitoring radioactive materials, including source and by-product materials.

However, an Operations Supervisor, who holds at least a limited Senior Reactor Operators license, is responsible for handling new and spent fuel.

In the case of fuel handling operations that alter the configuration of the reactor core, supervisory personnel with either a limited Senior Reactor Operator (SRO) or SRO license are directly responsible for movement of the fuel.

In the case of fuel handling operations that do not alter the configuration of the reactor core, qualified management personnel (such as Reactor Services or others designated by the Operations Department), who report to an Operations Supervisor, are directly responsible for movement of the fuel.

12.1.1.3 Administration Concerns The Byron/Braidwood administrative personnel have a considerable amount of experience, which was accumulated at operating stations. The health physics program is based on regulations and experience which includes or considers the following:

a. Detailed procedures are prepared and approved for radiation protection prior to reactor plant operation. Those procedures are a part of the station health physics program.

b. All incoming and outgoing shipments which may contain radioactive material are surveyed to assure compliance with 10 CFR 71, 10 CFR 73, and 49 CFR 100-180.

c. Radiological incidents are investigated and documented in order to minimize the potential for recurrence. Reports are made to the NRC in accordance with 10 CFR 20.

d. Periodic radiation, contamination, and airborne activity surveys are performed and recorded to document radiological conditions. Records of the surveys are maintained in accordance with 10 CFR 20.

12.1-2 REVISION 17 - DECEMBER 2018

B/B-UFSAR

e. Records of occupational radiation exposure are maintained and reports are made to the NRC as required by 10 CFR 20, and to individuals as required by 10 CFR 19.13.

12.1-2a REVISION 8 - DECEMBER 2000

B/B-UFSAR

f. Posted areas are segregated and identified in accordance with 10 CFR 20. A combination of administrative controls and physical barriers are utilized to control access to high and very high radiation areas in accordance with 10CFR20 and the Technical Specifications.

g. Personnel are provided with personnel radiation monitoring equipment to measure their radiation exposure in accordance with 10 CFR 20.

h. Process radiation, area radiation, portable radiation, and airborne radioactivity monitoring instrumentation are periodically calibrated as required.

i. Access control points are established to separate potentially contaminated areas from uncontaminated areas of the station.

j. Protective clothing is used as required to help prevent personnel contamination and the spread of contamination from one area to another.

k. Tools and equipment used in radiological posted areas are surveyed for contamination before removal to an uncontrolled area. Contaminated tools and equipment removed from a contaminated area are packaged as necessary to prevent the spread of contamination to uncontrolled areas.

l. Radiation work permits (RWP) are issued for certain jobs in accordance with the station radiation protection procedures. Jobs involving significant radiation exposure to personnel are preplanned.

(Where conditions dictate a mock-up is used for practice to reduce exposure time on the actual job.

The use of special tools and temporary shielding to reduce personnel exposure is evaluated on a job-by-job basis.)

m. A bioassay program is included as part of the health physics program. This program includes whole body counting and/or a urinalysis sampling program to measure the uptake of radioactive material.

n. An environmental radiological monitoring program is in operation to measure any effect of the station on the surrounding environment.

o. All significant radioactive effluent pathways from the station are monitored and records maintained.

12.1-3 REVISION 8 - DECEMBER 2000

B/B-UFSAR

p. There are no special lighting requirements for high radiation areas.

q. Known radiation sources are marked or identified as such in efforts to reduce personnel time in regions of the exposure field and increase personnel distance from the source of exposure. "Hot-spot" labels are utilized on some localized radiation sources as deemed appropriate.

12.1.2 Design Considerations Careful design can contribute greatly to the reduction of occupational radiation exposures. Radiation protection design considerations include shielding radioactive components, reducing the need for maintenance, enhancing the accessibility of equipment, reducing the source strength relative to personnel through remote handling, minimizing leakage and streaming, providing adequate ventilation, and preflushing contaminated systems.

Byron/Braidwood radiation protection design considerations establish a practical basis for maintaining radiation exposures ALARA. The direction is established by a set of radiation protection design goals. Conservatively set criteria in facility and equipment design, experience from past designs and operating plants incorporated to improve the present design, and mechanisms established for design review, are implemented to fulfill the ALARA requirement. (Radiation protection design features which are provided to maintain personnel radiation exposures ALARA are described in Section 12.3.)

12.1.2.1 Radiation Protection Design Goals Byron/Braidwood radiation protection design goals are directed to ensure compliance with the standards for radiation protection specified in 10 CFR 20. The following sequence of design goals was used as a basis for maintaining radiation exposures as low as is reasonably achievable.

a. Establish design dose rates for general access areas based upon Commonwealth Edison's experience and 10 CFR 20 regulations.

b. Determine the most severe mode of operation for each piece of equipment and section of pipe (Section 12.2).

c. Based upon source terms, determine the source for each piece of equipment or pipe (Section 12.2).

d. Determine shielding required to maintain design dose rates.

e. Determine advantages and disadvantages of equipment locations, orientation, and segregation.

12.1-4

B/B-UFSAR

f. Use predetermined guidelines and criteria for locating piping and penetrations (Section 12.3).

g. Make changes in design wherever practicable to achieve ALARA exposures.

12.1.2.2 Facility Design Considerations Byron/Braidwood's radiation protection design goals are expanded to design objectives. These objectives are categorized into several radiation protection concerns, which are described in the following subsections. Station layout considers direct radiation (for this section, direct radiation is defined as scattered and unscattered gamma and/or neutron rays from a [several]

nonairborne radiation source(s)), and ventilation considers airborne radioactivity (see Subsection 12.2.2.3). Health physics and access control are concerned with both direct and airborne radioactivity. Control of radioactive fluids and effluents is concerned with the processing and detection of radioactive materials. The assumptions of primary coolant activity listed according to isotope are given in Table 12.2-2. The majority of the other source terms were developed, from these activities.

The design objectives are coupled with operating experience to obtain an improved station design.

12.1.2.2.1 Station Layout (Shielding)

The shielding was arranged and designed to the following objectives:

a. A sufficient quantity of access paths (general access areas) are furnished to allow personnel access to equipment.

b. The radiation levels in general access areas are to be kept ALARA.

c. Sufficient shielding is provided to control the amount of direct radiation present in a general access area.

d. Radiation areas are classified into zones according to expected (maximum) radiation levels.

e. Segregation of radiation zones is employed whenever practicable.

f. Shielding must accommodate equipment removal and maintenance.

g. Radiation "hot spots" are to be expected along the face of some shielding walls due to penetration and 12.1-5

B/B-UFSAR embedded system piping (i.e., nonradioactive piping designed for the passage of air, steam, water, or oil). A radiation "hot spot" is a small area that has a higher dose rate than the surrounding areas. A "hot spot" has a set maximum value that is based upon the adjacent design dose rates.

h. The radiation protection design is to be based upon the design criteria given in Section 12.3.

12.1.2.2.2 Ventilation The station ventilation systems aid in heat removal and control of airborne radioactivity. Ventilation systems are designed to direct potentially airborne radioactive material from occupied areas towards the station vent stack. The remaining HVAC systems have special functions (e.g., laboratory hood exhaust). The ventilation systems are described in greater detail in Section 9.4. The radiation protection aspects of the systems are discussed in Subsection 12.3.3.

12.1.2.2.3 Health Physics The radiation protection design objectives for health physics are:

a. The station's radiation protection monitoring equipment is located (and is of sufficient quality) to detect excessive airborne radioactivity and high radiation levels.

b. Personnel radiation monitoring equipment is required to measure and record personnel radiation exposure.

c. Radioactive effluent release paths to the environment are monitored.

d. Facilities for analysis of radioactive samples are furnished.

e. Cleaning and decontamination facilities are provided for equipment and protective clothing.

f. Periodic radiation surveys are performed when required, such as for maintenance in radiation areas, receiving or shipping radioactive material, and decontamination and maintenance of equipment, parts, and tools.

12.1.2.2.4 Access Control Access to radioactive equipment is designed so that with properly trained personnel, radiation exposures during all modes of station operation meet the ALARA requirements. Access to radiation areas is strictly controlled.

12.1-6

B/B-UFSAR 12.1.2.2.5 Control of Radioactive Fluids and Effluents Radioactive fluids (liquids and gases) are contained and controlled to keep the release of radioactive materials to general access areas and the environment ALARA. This objective applies to drain liquids, airbornes, and process liquids and gases (e.g., reactor water, fuel pool water, radwaste water, and off-gas). The number of release paths is minimized in order to simplify control.

12.1.2.2.6 Safety Objectives

a. The 10 CFR 20 limits are maintained for operating personnel and the general public.

b. The 10 CFR 50 limits for the control room are met for a design-basis accident (DBA) and lesser accidents.

c. Radiation protection design objectives related to 10 CFR 100 for accidents analyzed using TID-14844 and 10 CFR 50.67 for AST are given in Chapter 15.0 12.1.2.3 Improvements in Facility Design Due to Past Experience and Operation At the time of the design and construction of Byron and Braidwood, Commonwealth Edison operated five licensed BWRs and two PWRs (see Chapter 1.0). The operating experience obtained from these stations has been incorporated into the design of Byron/ Braidwood Stations. In addition, published information on radiation problems and radiation protection (in nuclear power stations) was used to anticipate and minimize occupational radiation exposure. Experienced operating personnel continually reviewed the station design as the design progressed, and provided recommendations based on their experience.

Routine survey data from Commonwealth Edison's operating stations has been used to correct or improve the design of Byron/

Braidwood Stations. Some design improvements directly attributed to experiences and operations are as follows; others are discussed in Section 12.3.

a. An adequate number of equipment decontamination areas have been included to reduce congestion and reduce maintenance time.

b. Concrete shield walls, floors, and ceiling are coated with a nonporous coating to enhance decontamination wherever a potential for leakage or spillage of radioactive material exists on these surfaces.

c. To the extent practicable, all valves servicing radioactive or potentially radioactive equipment are centrally located in shielded valve aisles apart from 12.1-7 REVISION 12 - DECEMBER 2008

B/B-UFSAR the equipment serviced; walk-in valve aisles are used. Where practicable, no valves are located in pipe tunnels.

d. All radioactive or potentially radioactive manually operated valves and associated piping are shielded from the valve operating area when practicable.

Remote manual valve operators connected to manually operated or geared handwheels extending through the shielding to the valve operating area are used (see Figure 12.3-3). Valve operating personnel are thus protected from radiation due to radioactivity in the valves and associated fluid piping in the valve aisle.

e. To reduce the amount of radioactive material in valve aisles, radioactive pipe runs to and from valves aisles are minimized by maximizing the amount of radioactive runs behind the shield wall placed between the piece of equipment and the valves.

f. Motor and pneumatic operated valves (generally higher maintenance items than manually operated valves) which are in radioactive or potentially radioactive service, are located in areas shielded from the component serviced by the valve. This minimizes personnel exposure during valve maintenance and inspection.

g. Valves servicing radioactive or potentially radioactive equipment are installed and positioned relative to other valves so as to minimize maintenance time. Space is provided around valves so that compensatory shielding (such as lead blankets) can be used as needed.

h. Components associated with control of the instrument air supply to air operated valves, are not themselves radioactive or potentially radioactive, and are located in low radiation areas.

i. Controls are installed in the lowest practicable radiation zone; use of transducers is maximized in high radiation areas. Instrument readouts are located in areas which will result in the lowest personnel exposure, if consistent with other requirements such as instrument accuracy and precision.

j. Instrument readouts are designed and located to minimize the time and exposure necessary to take a reading. They are positioned in readily accessible, adequately lighted areas, at a convenient elevation for observation and parallax correction. They must face in a direction convenient for reading, have 12.1-8

B/B-UFSAR easily readable numbers and pointers; the application of scale multipliers is minimized.

k. Shielding separates pumps from their associated tanks or other vessels.

1. Space and adequate floor strength for temporary shielding is supplied where practicable.

More examples of how Commonwealth Edison's experience has contributed to the Byron/ Braidwood Stations design can be found in Section 12.3.

12.1.2.4 Equipment Design Considerations Radiation protection design consideration of equipment involves shielding, equipment access, equipment selection, and equipment maintenance. Equipment design objectives deal with access to, and segregation of, radioactive equipment. The following are the equipment design objectives for radiation protection:

a. Equipment which processes fluids with low radioactivity are located in separate cubicles from equipment which processes highly radioactive fluids.

b. Galleries, hatches, and gratings are provided as needed to allow access to equipment from the top, especially if the piece of equipment is high above a floor.

c. Equipment is located in accessible parts of cubicles. Equipment frequently changed in whole or in part is readily accessible.

d. Cranes or lifting lugs are provided as needed for equipment servicing, maintenance, and removal.

e. Localized shielding or space and adequate structure for localized shielding is provided as part of the shielding design.

f. Unmortared removable block walls or easily removable floor or wall plugs are provided to minimize the radiation exposure in gaining access to highly radioactive components when removal (e.g., tube pulling) is required.

12.1.2.5 Equipment Selection The selection of equipment to handle and process radioactive materials is based upon system requirements and radiation protection requirements. Consideration is given to minimizing leakage, spillage, and maintenance requirements. Material and 12.1-9

B/B-UFSAR coating selection are chosen for decontamination properties as well as durability. Some components which may become contaminated are designed with provisions for flushing or cleaning. Reduced occupational radiation exposure is attained by utilizing operating experience and where practical, providing prudent equipment selections such as:

a. plug valves which require less maintenance in place of diaphragm valves;

b. diaphragm seal valves which require no packing;

c. longer-life graphite-filled packing, instead of standard packing;

d. fluid connections for the capability to back flush;

e. remote systems (or connections) for remote chemical cleaning where practicable;

f. air connections to tanks containing spargers to allow for air injection to uncake contaminates;

g. cross-ties between redundant equipment and/or related equipment capable of redundant operation to allow removal of contaminated equipment from service;

h. pumps with flanged connections to allow quick removal and installation;

i. mechanical seal flushing lines on pumps to reduce the accumulation of radioactive material in the seals;

j. remote filter handling equipment for radwaste disposal; and

k. drains on tanks flush with inside surface of the tanks.

12.1.2.6 Overall Impact of Design Considerations Special attention has been given, as noted above and elsewhere in this chapter, to maintaining occupational radiation dose ALARA -

while establishing the final design of Byron/Braidwood Stations.

The design of facilities, equipment, structures, and access areas consider exposure obtained during routine operations (sampling, surveys, inspections, etc.), transient operations (changing power levels, startup, and shutdown), operational occurrences (identification, removal from service, etc.), maintenance, moving and storing radioactive materials, and accidents. These designs take into account equipment removal, decontamination, ventilation, orientation of equipment, in situ calibration and maintenance, sampling, monitoring, shielding, controlling contaminated fluids, minimizing leakage and spillage, and radiation exposure.

12.1-10 REVISION 1 - DECEMBER 1989

B/B-UFSAR The station staff includes health physicists as described in Subsection 13.1.1.3. Experience in radiation protection has been incorporated into the design of Byron/Braidwood during review and comment stages. In addition, design reviews have been conducted by other competent health physicists.

The design philosophy established for Byron/Braidwood strives to maintain occupational radiation exposure ALARA and is in compliance with applicable regulations.

12.1.2.7 Radiation Protection Design Review

a. Reviewers of the Radiation Protection Design The station owner has the responsibility for the radiation protection design review on the Byron and Braidwood Stations. Commonwealth Edison utilized Westinghouse and Sargent & Lundy to review the Byron/Braidwood Stations' radiation protection design.

Westinghouse employs system analysis engineers, competent in the area of health physics and radiation protection, to work with system design engineers.

Although many groups within the Westinghouse Systems Division (SD) are available when required, the two major sections responsible for radiation protection review are Plant and Systems Evaluation Licensing, within the Nuclear Safety Department, and Radiation and Systems Analysis within the Engineering Department. The managers of these two sections report through the management of their respective departments to the SD General Manager, who is responsible for the overall design of RESAR-414 plants.

The A-E, Sargent & Lundy, performs ALARA Radiation Protection Design Reviews at key points in the balance of plant design. These reviews are independent of the owner's reviews and incorporate the instructions of the owner. The radiation protection design reviews conducted by Sargent & Lundy, cover access control, radiation shielding, radiation monitoring, radiation protection facilities, and control of airborne contamination in accordance with the ALARA concepts in Sections C.2 and C.4 of Regulatory Guide 8.8. The Sargent & Lundy ALARA review is conducted according to written procedures which establish a review committee and a committee chairperson. The chairperson is an experienced radiation protection specialist and is responsible for the design review; he assigns committee members and additional reviewers as necessary to review tasks in their area of expertise. The review committee issues 12.1-11

B/B-UFSAR a report summarizing its review and its conclusions.

A summary of the qualifications of the personnel who participated in the most recent Sargent & Lundy ALARA Radiation Protection Design Review are given in Table 12.1-2. The review team consisted of the committee chairperson, at least three committee members and two additional reviewers.

Types of personnel that have been involved in the radiation protection review are given in Tables 12.1-1 and 12.1-2.

b. Recordkeeping of the Radiation Protection Design Review Process Design information is logged and sent to the owner for comments. Portions of the design information involve radiation shielding, monitoring, laboratory facilities and other radiation considerations. These items are directed to the responsible radiation protection reviewer. Comments are sent through both project manager's divisions (owner and designer).

Radiation protection comments and requested changes are forwarded to the engineer responsible for the radiation protection (RP) design. The RP designer responds to the comments and requests. He then files the comments, requests, and the response. The RP designer makes the required design changes. The project management divisions coordinate and document the changes.

The personnel with expertise in radiation protection within the groups stated above participate in the design review process in a systematic manner. The procedures to assure radiation protection functions needed to prevent or mitigate consequences of postulated accidents that could cause undue risk to the health and safety of the public are formally documented.

The NRC has reviewed the Westinghouse policy, design, and operational considerations related to assuring that occupational radiation exposures are ALARA for the RESAR-3S and RESAR-414 designs. They have concluded that Westinghouse has shown sufficient concern and familiarity with the ALARA principles in the areas of design considerations such that this aspect of radiation protection is acceptable. There are no substantial differences between RESAR-414, RESAR-3S and the Byron/Braidwood design in those areas that affect ALARA.

12.1-12

B/B-UFSAR

c. Radiation Protection Design Techniques to Reduce Person-Rem Exposure

1. The utilization of removable unmortared block wall sections (instead of mortared sections) for some equipment significantly reduces the number of person-hours spent in radiation areas.

2. Probe holes were placed in most removable hatches of filter and demineralizer cubicles. These holes allow radiation monitoring of the cubicles prior to removing its hatch. The radiation data from the monitor allows radiation protection personnel better control of occupational exposure.

3. Area radiation monitors (ARMs) were placed in valve aisles which serve two or more highly radioactive systems. These ARMs provide a warning on high-radiation and help to prevent high levels of unexpected exposure from the startup of an inactive system while performing maintenance on another system.

12.1.3 Operational Considerations Operational radiation protection objectives deal with access to radiation areas, exposure to personnel, and decontamination.

Working at or near highly radioactive components requires planning, special methods, and criteria directed toward keeping occupational radiation exposure ALARA. Job training and briefing for selected high exposure jobs contribute toward reduced exposures. Decontamination also helps to reduce exposure.

Procedures and techniques are based upon operational criteria and experience. Procedures are discussed in Section 13.5.

12.1.3.1 Operational Objectives The operational radiation protection objectives include the following:

a. knowledge of station design;

b. experienced personnel to direct and train other personnel;

c. detailed job planning and pre-job meetings for high exposure work;

d. job simulations to improve productivity on the job, thereby keeping exposure ALARA;

e. briefings after selected high exposure jobs to identify time consuming work and to identify problems; 12.1-13 REVISION 7 - DECEMBER 1998

B/B-UFSAR

f. improving procedures and techniques (defined in the following subsection) for future jobs;

g. use of radiation monitoring equipment to detect airborne radioactivity concentrations and high radiation levels and to measure and record personnel radiation exposure;

h. analysis of radioactive samples to monitor chemistry, check for radiation release, etc.;

i. use of cleaning and decontamination facilities for equipment and protective clothing; and

j. use of periodic radiation surveys are required.

12.1.3.2 Implementation of Procedures and Techniques The criteria or conditions under which various operating procedures and techniques for ensuring that occupational radiation exposures are ALARA for systems associated with radioactive liquids, gases, and solids, along with the means for planning and developing procedures for radiation exposure-related operations, are given in the following:

a. Section 12.1, "Ensuring That Occupational Radiation Exposures Are ALARA;"

b. Section 12.3, "Radiation Protection Design Features;"

and

c. Section 12.5, "Health Physics Program."

12.1.3.3 Implementation of Exposure Tracking and Exposure Reduction Program The Exelon Generation Company commitment to the ALARA principle is discussed in Subsection 12.1.1. The use of radiation work permits is discussed in Subsection 12.1.1.3.

Self-reading dosimeters are used at Byron/Braidwood stations to record estimates of daily exposures received by each individual worker. This information enables the Radiation Protection Department to spot significant individual exposures prior to processing other monitoring dosimetry. Work group person-rem summaries are generated by a computerized dose tracking program.

The summaries serve to alert the station health physics staff and the corporate office of the trends in person-rem expenditures.

Commonwealth Edison began a Radiation Evaluation Program (REP) in April of 1976. REP is a computer based occupational dose accounting system used to document, by work group, the dose 12.1-14 REVISION 8 - DECEMBER 2000

B/B-UFSAR expenditure resulting from work performed on various plant systems and components. In addition to each work group's dose and the plant component worked on, the program documents the total work effort in person-hours and include a brief description of the work performed.

The REP program applications are:

a. To provide timely radiological feedback information to the engineering and production departments and architect-engineer consultants for consideration in new plant design and to enable corrective action to be taken at existing stations.

b. To identify and compile dose histories on specific sources of occupational dose that might be reduced through improved station working and shielding procedures and training programs.

c. To provide data for comparison studies of specific sources of occupational exposure among similar Exelon Generation Company nuclear stations with relevant factors such as reactor equipment and plant layout, etc., taken into account.

d. To demonstrate an "active ALARA program."

The station has an ALARA Review Committee. This committee is composed of the manager of each affected department, the station manager, and Radiation Protection Department personnel. The charter of the committee is to advise the station manager on ALARA matters. The committee reviews annual exposure reduction goals as a part of its activities. The committee meets periodically as stated in subsection 12.1.1.1.

12.1-15 REVISION 8 - DECEMBER 2000

B/B-UFSAR TABLE 12.1-1 NSSS RADIATION PROTECTION PERSONNEL RADIATION PROTECTION JOB TITLES REVIEW RESPONSIBILITIES EDUCATION EXPERIENCE Manager of Interfaces between the BS or higher 5 years as a Energy and Engineering Department in engineer- lead engineer Environmental and the NRC. He ing or the or manager.

Analysis reviews, coordinates, physical Background in and supplies input sciences nuclear and for Chapters 1, 2, chemical en-11, 12, and 15 of vironmental the Safety Analysis engineering Reports.

Manager of Provides radiation MS or equi- 6 years expe-Radiation protection guidance. valent in rience in and System Analyzes plant ra- mechanical, nuclear plant Analysis diation sources and nuclear, or system opera-exposure from and chemical tion or to components. engineering design Occupational ra-diation exposure design review.

12.1-16

B/B-UFSAR TABLE 12.1-2 RADIATION PROTECTION PERSONNEL PARTICIPATING IN THE AEs SUBMITTAL RADIATION PROTECTION DESIGN REVIEW EDUCATION OF EXPERIENCE OF*

SPECIFIC SPECIFIC JOB TITLE RESPONSIBILITIES REVIEWERS REVIEWERS Chairperson Coordinate review Chairperson: Over 25 years NSLD Radia- by the committee experience in tion Protec- B.S.E.E. the nuclear tion Design Assign reviewers industry and Review Certified with the AEC.

Committee Assign review tasks Health Physicist Resolve disputes Registered Approve committee Professional conclusions Engineer Terminate review Committee Assigned a Members:

Members specific area of responsibility Ph.D., NE Over 7 years in nuclear Summarize review engineering and responses radiation engineering Make recommenda-tions and Ph.D., Health One year in appraisals of Physics health physics plant's RP design MS, NE Over 13 years Registered in nuclear Profes- engineering, sional radiation Engineer engineering, and health physics Experience at time of design review.

12.1-17 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.1-2 (Cont'd)

EDUCATION OF EXPERIENCE OF*

SPECIFIC SPECIFIC JOB TITLE RESPONSIBILITIES REVIEWERS REVIEWERS Reviewers Assigned a Reviewers:

(In addition specific area of to committee responsibility Ph.D., NE 4 years in members) nuclear Review completeness engineering of station's radi- and radiation ation protection engineering Design.

MS, NE 3 years in Identify nuclear deficiencies engineering and health Make physics recommendations 12.1-18 REVISION 1 - DECEMBER 1989

B/B-UFSAR 12.2 RADIATION SOURCES 12.2.1 Contained Sources The sources given in the subsection are based on the following parameters:

a. power = 3565 MW;

b. operation with defects in cladding or rods generating 1% of the core rated power;

c. reactor coolant mass = 2.42 x 108 grams; and

d. reactor coolant purification rate = 75 gpm at 130F.

The design-basis shielding sources are more detailed and conservative than the realistic sources presented in Subsection 11.1.2. The conservative (design-basis) source model is described in Subsection 11.1.2.1.

The impact of a core power uprate on the design-basis shielding radiation source terms discussed above is provided in Section 12.2.4. The original licensed power level was 3411 MWt. The original source term and shielding analyses were performed at a power level of 3565 MWt. Byron and Braidwood Nuclear stations have uprated the core power level twice. First to a core power level of 3586.6 MWt, then to the Measurement Uncertainty Recapture uprate power level of 3645 MWt.

12.2.1.1 NSSS Sources Reactor Coolant Concentrations of activation products in the reactor water are given in Table 12.2-1 for Nitrogen-16 and in Table 12.2-2 for the other activation products and fission products.

Steam Generator The activities on the primary side surfaces of the steam generator are used in determining access limitations in and around the steam generators at plant shutdown. Nominal values of deposited activity are listed in Table 12.2-3 for several operating times.

Pressurizer The radioactive sources in the pressurizer, steam, and liquid phases, as well as the deposited sources are tabulated in Tables 12.2-4, 12.2-5, and 12.2-6 respectively.

Letdown Coolant Fission and Corrosion Products The spectral source strengths in the purification letdown flow are tabulated in Table 12.2-7. The sources assume sufficient delay time from the reactor coolant loop for decay of the N-16 isotope and a fluid temperature of approximately 130F (i.e.,

downstream of the letdown heat exchanger).

12.2-1 REVISION 15 - DECEMBER 2014

B/B-UFSAR Volume Control Tank The sources in the volume control tank are itemized in Table 12.2-8. These sources correspond to a nominal operating level in 12.2-1a REVISION 9 - DECEMBER 2002

B/B-UFSAR the tank of 160 ft3 in the liquid phase and 240 ft3 in the vapor phase.

Recycle Holdup Tank The radiation sources in the recycle holdup tank exist in both the vapor and liquid phases. The vapor sources are used to determine the holdup tank shielding requirements, whereas the liquid sources are the basis for maximum evaporator activities.

For the vapor space inventory, it is assumed that all gases in the water entering the tank flashes in the vapor space. The Kr-88 isotope is the major isotope in terms of shielding requirements. The vapor sources are based on the time when the Kr-88 inventory is a maximum in the tank. The vapor sources are listed in Table 12.2-9.

The liquid phase activities are based on the assumption that all the gases remain in solution. The basis for this assumption is that the holdup tank liquid serves as feedwater to the recycle evaporator. The liquid phase sources are listed in Table 12.2-10.

Recycle Evaporator The sources associated with the recycle evaporator package are specified in Table 12.2-11. In the package, gaseous activity is concentrated in the vent condenser portion while particulate and dissolved activity is concentrated in the evaporator section.

Residual Heat Removal Loop The maximum specific source strengths in the residual heat removal loop are tabulated in Table 12.2-12. The residual heat removal loop is placed in operation approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reactor shutdown and reduces the reactor coolant temperature to approximately 120F within about 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after shutdown. The sources are maximum values with credit taken for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of activity decay and purification.

Ion Exchangers, Chemical Volume Control System (CVCS)

The radiation sources in the ion exchangers of the CVCS are tabulated in Tables 12.2-13 through 12.2-16. The mixed bed retains the fission product activity, both cations and anions, and the corrosion product (crud) metals. The cation bed can be used intermittently to remove lithium for pH control, and supplements the mixed bed in removing Y, Cs, Mo, and the crud metals.

The boron thermal regeneration beds are used to regulate the boron concentration in the reactor coolant water. They are utilized during load follow operations, and in removing boron from the coolant as the nuclear fuel is depleted. These 12.2-2

B/B-UFSAR demineralizers also collect radioactive anions, such as iodine, which may have passed through the mixed bed.

Recycle Evaporator Condensate Demineralizer Sources for the recycle evaporator condensate demineralizer are given in Table 12.2-17.

Filters Sources for the reactor coolant filter, seal water return filter, recycle evaporator feed filter, spent fuel pit filter, spent fuel pit skimmer filter, seal water injection filter, recycle evaporator concentrates filter, and recycle evaporator condensate filter, are given in Tables 12.2-19 through 12.2-22.

Reactor Core The core gamma sources (after shutdown) are used to establish radiation shielding requirements during refueling operations and during shipment of spent fuel. The sources associated with the spent fuel are based on an average power assembly with an irradiation time of 108 seconds (3.1 years). These source strengths per unit volume of homogenized core are tabulated in Table 12.2-23 for various times after shutdown.

Irradiated Control Rods The irradiated control rod sources are used in establishing radiation shielding requirements during refueling operations and during shipping of irradiated rods. The absorber material used in the control rods is silver-indium-cadmium (Ag-In-Cd) or hafnium.

The source strengths associated with the control rods are listed in Tables 12.2-24 and 12.2-24a for various times after shutdown.

The values are per cm of height of a single rod for an irradiation period of 100,000 hours0 days 0 hours 0 weeks 0 months .

Refueling Water Prior to refueling, the radioactivity in the reactor coolant is reduced by operating the purification system at the maximum letdown rate. Particulate (soluble) activity such as cesium, iodine, and the metals are removed by the mixed bed and cation bed demineralizers. Radioactive gases are removed by the volume control tank.

Sources for the spent fuel pit demineralizer are given in Table 12.2-18.

Based on a direct exposure dose rate of 2.5 mrem/hr at the surface of the refueling water, the accumulative isotopic concentration in the refueling water (after dilution) should not 12.2-3

B/B-UFSAR exceed one maximum permissible concentration (< MPC), as defined by 10 CFR 20, Appendix B, Table II, column 2 (the term MPC as used in this section refers to a 10CFR20 limit in effect prior to January 1, 1994) and Equation 12.2-1. The dilution consists of the complete mixing of the reactor coolant in the vessel with stored refueling water, approximately 1 to 10, respectively. The resulting mixture must satisfy the following relationship:

CA CB CC DD

+ + + + ... 1 (12.2-1)

MPCA MPCB MPCC MPCD where the subscript identifies the isotope, C is the concentration in Ci/cc and MPC is one maximum permissible concentration in Ci/cc.

Table 12.2-25 gives the maximum allowable concentration (same as MPC) for some dominant isotopes.

The refueling pool purification system is designed to maintain the relationship given in Equation 12.2-1.

Irradiated Incore Detectors Table 12.2-26 shows the incore fission chamber sources energy spectrum after three months of irradiation and one day decay.

The incore detector drive wire sources are used in establishing the radiation shielding requirements for the wires when the detectors are not in use and during shipment when the detectors have failed.

Table 12.2-27 lists the detector drive wire sources per cm of length of wire, assuming that the detector has been placed in the core for 1 year.

Process Piping The radiation sources in process piping are derived from the activity in the process fluid plus an estimate of crud buildup.

The concentrations in the process fluids are given in the following:

Tables 12.2-1, 12.2-2, 12.2-4, 12.2-5, 12.2-8 through 12.2-18, 12.2-25, 12.2-29, 12.2-30, 12.2-31, and 12.2-33.

The crud buildup levels are estimated using data from operating stations (References 3 and 4).

Gas Decay Tanks The gas decay tank inventory used in the Chapter 15.0 accident analysis reaches a maximum while degassing the reactor coolant 12.2-4 REVISION 8 - DECEMBER 2000

B/B-UFSAR system during a cold shutdown. The gases removed by the volume control tanks are vented to the waste gas system. The maximum activity for the shielding design of a single gas decay tank is shown in Table 12.2-28, and represents the inventory present at the time Kr-85 is at a maximum. The shield thickness of the gas decay tank cubicle was determined using the maximum activity in the two tanks.

Spent Fuel Pit The sources for the spent fuel pit fuel handling accident are given in Table 12.2-29.

12.2.1.2 Balance of Plant Shielding Design-Basis Sources Shielding source terms are supplied here for components which contain liquid from processed reactor coolant and which are considered part of the radioactive waste management system.

Sources are presented in Tables 12.2-30 and 12.2-31 for the station drains and steam generator blowdown which are input sources to the liquid radwaste processing system. In addition, sources are presented for the spent resin storage tank in Table 12.2-43. The primary assumptions used in the calculation of source terms are presented in the following sections.

12.2.1.2.1 Blowdown Sources Blowdown sources are based on 1 gpm primary-to-secondary tube leakage for one steam generator for 14 days. The total blowdown rate for four steam generators during reactor coolant leakage is approximately 135 gpm (see Subsection 10.4.8), so input into the blowdown stream is 1/135 times primary coolant activity.

12.2.1.2.2 Radwaste Processing Sources Liquid radwaste processing sources are based on the radioactive sources contained in the drain tanks shown in Tables 12.2-30 and 12.2-31.

Turbine building floor drain tanks are based on main steam condensate activity which consists of blowdown sources (1/135 times primary coolant activity) multiplied by the steam generator partition factor (10-2 for iodines and 10-3 for noniodines),

divided by 2 since the turbine building services two units.

The turbine building drains source bases along with the chemical drain tank, chemical/regeneration waste drain tank, auxiliary building floor and equipment drain tanks, and laundry drain tank source bases are shown on Table 12.2-32 as fractions of primary coolant obtained from Reference 1. Laundry drain tank sources are further adjusted by incorporating laundry drain sources measured at the Zion and Quad-Cities Stations. These sources are shown on Table 12.2-33.

12.2-5 REVISION 7 - DECEMBER 1998

BYRON-UFSAR The decontamination factors (DFs) used for filters, demineralizers, and evaporators are given in Table 12.2-34 for the elements found in the sources. Data is from References 1 and

2. The sources for the components of the radwaste processing system shown on Drawing M-48a are given in Tables 12.2-35 through 12.2-41. Included are shielding design-basis source terms for the blowdown mixed bed demineralizer, radwaste mixed bed demineralizer, concentrates holding tank, blowdown prefilter and afterfilter, radwaste afterfilter, turbine building equipment drain filter, turbine building floor drain filter, auxiliary building equipment drain filter, regeneration waste drain filter, chemical drain filter, laundry drain filter, radwaste evaporator, 30,000 gallon release tank, laundry drain tank, blowdown monitor tank, radwaste evaporator monitor tank, and the auxiliary building floor drain filter. Source terms for the steam generator blowdown prefilters were calculated using the volume of equipment 1/2WX02MA,B (housing-only prefilter vessels).

The design-basis operating time for radwaste processing equipment was taken as 30 days which is sufficient time for most radionuclides of concern to build up to equilibrium.

12.2.1.2.3 Spent Resin Storage Tank The spent resin storage tank inventory assumes that the demineralizers with the highest potential activities will send their resins to the tank. The demineralizers considered and the assumed fractional contribution of each is shown in Table 12.2-42. The demineralizers are assumed to operate for one core cycle at full power and 1% failed fuel for both units.

Demineralizer sources are given in Subsection 12.2.1.1. The spent resin storage tank radionuclide inventory is given in Table 12.2-43.

12.2.1.2.4 Radwaste Solidification System The shielding source terms for the decanting and drumming equipment storage area are based on the drumming of wastes from the spent resin storage tank shown in Table 12.2-43, column 2, with no decay time assumed. Spent resins are the highest activity sources to be handled by the radwaste solidification system and result in a conservative shielding design basis.

The composition of a design-basis radwaste drum used in the shielding of the radwaste drum storage areas is shown in Table 12.2-44. The sources used for intermediate activity drum storage area shielding are spent resins decayed for 90 days, which assumes that resins are stored for this average time period in the spent resin storage tank prior to drumming. These sources result in a dose rate of 80 R/hr for a single drum at contact.

Spent resin storage tank sources decayed for 90 days are given in Table 12.2-43.

The sources used for the low activity drum storage area shielding are radwaste evaporator concentrates shown in Table 12.2-39.

12.2-6 REVISION 9 - DECEMBER 2002

BRAIDWOOD-UFSAR The decontamination factors (DFs) used for filters, demineralizers, and evaporators are given in Table 12.2-34 for the elements found in the sources. Data is from References 1 and

2. The sources for the components of the radwaste processing system shown on Drawing M-48A are given in Tables 12.2-35 through 12.2-41. Included are shielding design-basis source terms for the blowdown mixed bed demineralizer, radwaste mixed bed demineralizer, low activity spent resin storage tank holding tank, blowdown prefilter and afterfilter, radwaste afterfilter, turbine building equipment drain filter, turbine building floor drain filter, auxiliary building equipment drain filter, regeneration waste drain filter, chemical drain filter, laundry drain filter, radwaste evaporator, 30,000 gallon release tank, laundry drain tank, blowdown monitor tank, radwaste evaporator monitor tank, and the auxiliary building floor drain filter.

Source terms for the steam generator blowdown prefilters were calculated using the volume of equipment 1/2WX02MA,B (housing-only prefilter vessels).

The design-basis operating time for radwaste processing equipment was taken as 30 days which is sufficient time for most radionuclides of concern to build up to equilibrium.

12.2.1.2.3 Spent Resin Storage Tank The high activity spent resin storage tank inventory assumes that the demineralizers with the highest potential activities will send their resins to the tank. The demineralizers considered and the assumed fractional contribution of each is shown in Table 12.2-42. The demineralizers are assumed to operate for one core cycle at full power and 1% failed fuel for both units.

Demineralizer sources are given in Subsection 12.2.1.1. The high activity spent resin storage tank radionuclide inventory is given in Table 12.2-43.

It is expected that only the resin from blowdown and radwaste mixed bed demineralizers will be sent to the low activity spent resin tank (formerly the concentrates holding tank).

The total amount of resin (Anion and Cation) per mixed bed for blowdown and radwaste systems is 113 ft3 and 29 ft3 respectively.

The demineralizers considered and the assumed fractional contribution of each is shown in Table 12.2-53.

The low activity spent resin tank expected radionuclide inventory is given in Table 12.2-54.

12.2-6a REVISION 9 - DECEMBER 2002

BRAIDWOOD-UFSAR 12.2.1.2.4 Radwaste Solidification System The shielding source terms for the decanting and drumming equipment storage area are based on the drumming of wastes from the spent resin storage tank shown in Table 12.2-43, column 2, with no decay time assumed. Spent resins are the highest activity sources to be handled by the radwaste solidification system and result in a conservative shielding design basis.

The composition of a design-basis radwaste drum used in the shielding of the radwaste drum storage areas is shown in Table 12.2-44. The sources used for intermediate activity drum storage area shielding are spent resins decayed for 90 days, which assumes that resins are stored for this average time period in the spent resin storage tank prior to drumming. These sources result in a dose rate of 80 R/hr for a single drum at contact.

Spent resin storage tank sources decayed for 90 days are given in Table 12.2-43.

The sources used for the low activity drum storage area shielding are radwaste evaporator concentrates shown in Table 12.2-39.

12.2-6b REVISION 5 - DECEMBER 1994

B/B-UFSAR These sources result in a dose rate of 3 R/hr at contact. The models used for shielding design are given in Subsection 12.3.2.1.9.

12.2.1.2.5 Volume Reduction System The volume reduction (VR) system is addressed in Section 11.4.

Source information and shielding design information has been intentionally deleted. Byron and Braidwood stations do not intend to use volume reduction system equipment.

12.2.1.3 Sources for HVAC Charcoal Filters Sources for the main auxiliary building charcoal filters and the off-gas vent charcoal filters are given in Table 12.2-45. The off-gas filter system has been modified such that all exhaust gases bypass the filter unit under all operating conditions. The following assumptions were made in the determination of the sources on the main auxiliary building charcoal filters.

a. Halogens become airborne only by evaporation from leaks of radioactive steam or water to the interior of the auxiliary building.

b. The steam or water contains reactor water sources.

c. The total leak rate is 20 gal/day for cold leakage and 1 gal/day for hot leakage. Partition factors are 0.001 and 0.1 for cold and hot leakage respectively.

d. Instantaneous complete mixing of the iodine throughout the volume of the auxiliary building occurs.

e. The exhaust rate through the filters is 20,000 ft3/min. Holdup time in the filters is 0.25 seconds.

12.2-7 REVISION 17 - DECEMBER 2018

B/B-UFSAR

f. There are six main auxiliary building charcoal filters with efficiencies of 95%.

The following assumptions were made in the determination of the sources on the off-gas charcoal filters. (The off-gas filter system has been modified such that all exhaust gases bypass the filter unit under all operating conditions.)

a. There is a primary to secondary leakage in one steam generator of 1 gpm with 1% failed fuel.

b. All four steam generators blow down at once. The blowdown rate for the one leaking steam generator is 90 gpm, while the blowdown rate for each of the three nonleaking steam generators is 15 gpm.

c. Partition factors are 0.01 for the steam generators, 5 x 10-4 for the main condenser, and 0.05 for the blowdown tank and condenser vents.

12.2-7a REVISION 17 - DECEMBER 2018

B/B-UFSAR

d. The main steam flow rate is 1.51 x 107 lb/hr.

e. The filter efficiency is 95%. Since the filters are in series, in the off-gas filter train, they may be considered together as one filter.

12.2.1.4 Old Steam Generator Storage Facility The old steam generator storage facility (OSGSF) is a reinforced concrete building that provides long-term storage and shielding for the four, old Unit 1 steam generators. The facility is located in the owner-controlled area outside of the security perimeter fence (see Figure 2.1-7 for Byron and Figure 2.1-4 for Braidwood). Shielding analysis for the OSGSF used measured dose rates obtained at each generator region in conjunction with waste samples to identify the dominant gamma-emitting isotopes (see Table 12.2-55).

12.2.2 Airborne Radioactive Material Sources With the exception of noble gases, sources of airborne radioactivity are generated from radioactive liquid sources by the mechanisms discussed in the following subsection. The generation of airborne radioactivity in radiation areas can affect the areas normally accessible to operating personnel, mainly pump and valve areas. The airborne radioactivity during normal operation for accessible areas is discussed in Subsection 12.2.2.3 The calculational model is given in Subsection 12.2.2.4.

In addition to affecting the station, the airborne radioactive material which exits in the filtration systems, enters the environment via the station vent stack.

12.2.2.1 Production of Radioactive Airborne Material Radioactive materials become airborne through evaporation and by being attached to suspended water droplets and water vapor. The water vapor comes from leaks in high energy lines (pressurized hot water). Suspended water droplets are created by sprays (usually leaks) and splashing. Evaporation occurs wherever there is standing water. Some examples are:

Component Airborne Method fuel pool evaporation radwaste evaporation (venting) high energy line leak vapor, evaporation 12.2-8 REVISION 7 - DECEMBER 1998

B/B-UFSAR Component Airborne Method spray from high energy vapor, droplets evaporation low energy line break evaporation spill droplets, evaporation The major contributors to airborne radioactivity during normal operation are: (1) leaks in the chemical volume control system, (2) evaporation from fuel pool, (3) leaks in radwaste systems, (4) venting of radwaste tanks, and (5) leaks in the charcoal-HEPA exhaust systems. Minor contributions are: (1) cleaning and decontaminating tools and equipment, (2) contaminated wearing apparel, and (3) sample preparation and analysis.

Some abnormal occurrences can cause airborne radioactivity; they are: (1) spills (i.e., overflows and splashing), (2) failure of a ventilation system, (3) cracks in piping, (4) failures of pump and valve seals, and (5) malfunctioning equipment.

12.2.2.2 Sources in Areas Normally Accessible to Operating Personnel Airborne radioactive material is expected to affect general access areas only during a ventilating system failure, or spillage of radioactive material in areas which are not sealed from general access areas. Airborne radioactive material is expected during refueling in maintenance areas, in labs (occasionally), and the hot instrument room.

The ventilation flow path is from areas of potentially low airborne radioactivity to ones of potentially high airborne radioactivity. The ventilation system has been designed to control the airborne radioactivity in the laboratories, maintenance areas, and the refueling floor of the reactor building. The concentration of airborne radioactivity is periodically determined by the radiation protection staff. The most significant radioactive isotopes are the halogens (primarily iodine). The iodines have the highest concentration in relation to the maximum permissible concentrations.

Maintenance accounts for a sizeable portion of the internal exposure of personnel because station personnel have to perform many of these functions in areas with relatively high airborne radioactivity. The airborne radioactivity is caused by leaks, spills, venting, etc. The airborne concentrations are calculated for the occurrences that are the most common, namely, leaks and venting. These concentrations are given in the next subsection.

Infrequent anticipated operational occurrences and abnormal occurrences are handled in the manner established in the personnel internal exposure program (Subsection 12.5.3.3).

12.2-9

B/B-UFSAR 12.2.2.3 Calculated Concentrations During Operation The calculated concentrations of airborne radioactive iodine in normally accessible cubicles are based upon the model given in Subsection 12.2.2.4. These concentrations are given in Tables 12.2-46, 12.2-47, and 12.2-48. The general access areas have very little if any airborne contaminants (i.e., <10-12 Ci/cc above background) during normal operation except for those mentioned in Subsection 12.2.2.2. Concentrations in normally accessible areas are determined by periodic air sampling as specified in the health physics program.

12.2.2.4 Models and Parameters Used in Calculations of Airborne Radioactivity Concentration For cubicles with non-radioactive supply air, the equation used to calculate the equilibrium airborne radioactivity concentrations during normal operation is as follows:

CA = [(L C P)/(7.48(V + F))] (12.2-2) where:

CA = airborne concentration in each cubicle (Ci/cc)

L = leak rate (gpm)

C = liquid concentration (Ci/cc)

P = fraction of activity released to air 7.48 = is conversion factor (7.48 gal/ft3)

= decay constant (min-1)

V = enclosed volume (ft3)

F = air exhaust flow rate (ft3/min).

12.2.2.5 Stack Effluents Ventilation system exhausts which are potentially radioactive are routed to the station vent stack. Each ventilation system is designed to exhaust into the station vent stack simultaneously with all other ventilation systems. Ventilation systems, containing potentially high airborne radioactivity concentrations are provided with filters specifically designed to hold-up or remove radioactive material (Section 9.4).

The dominating radioisotopes released through the station vent stack are the noble gases from the off-gas system and the vent filter system. The expected yearly releases during normal operation are discussed in Section 11.3.

12.2.3 Changes to Source Data Since PSAR Airborne radioactive material sources were not specified in the Byron/Braidwood PSAR. The entire Subsection 12.2.2 has been 12.2-10 REVISION 1 - DECEMBER 1989

B/B-UFSAR added. Sources for reactor coolant have been updated by Westinghouse. However, there have been no changes necessary in shielding requirements as a result of these revisions.

12.2.4 Impact of Uprate on Radiation Source Terms The original licensed power level was 3411 MWt. The original source terms for shielding analyses were performed at a power level of 3565 MWt. Byron and Braidwood Nuclear stations have uprated the core power level twice. First to a core power level of 3586.6 MWt, then to the Measurement Uncertainty Recapture uprate power level of 3645 MWt.

Core uprate will result in an approximate 0.6% increase in the inventory of core fission and activation products addressed in the original design basis source term/shielding analyses. The reactor coolant N-16 activity given in Table 12.2-1, the shutdown reactor core gamma sources presented in Table 12.2-23, the irradiated control rod sources in Tables 12.2-24 and 12.2-24a, the irradiated incore detector sources in Tables 12.2-26 and 12.2-27, and the spent fuel pit water activity for a fuel handling accident given in Table 12.2-29 will all increase by approximately 0.6% after core uprate. This small percentage increase is well within the uncertainty of the calculated design basis shielding source terms presented in this section (taking into consideration the accuracy of nuclear data, and the conservatism present in computation model simplification utilized in the original analyses). Consequently, these tables remain valid for uprate.

The deposited corrosion product activities on the primary side surfaces of the steam generators listed in Table 12.2-3 and on the pressurizer surface listed in Table 12.2-6 are based on the experience of large operating PWRs and are applicable for uprate.

Radiation sources in filters given in Tables 12.2-19 through 12.2-22 are conservative maximum values based on operating experience and remain valid for uprate.

The original design basis RCS activity given in Table 12.2-2 (same as Table 11.1-2) is re-calculated at a core power level of 3658.3 MWt which is 2% above the uprated power level. A more realistic nominal coolant mass of 2.477E8 gms is utilized in lieu of the conservative low RCS water mass of 2.42E8 gms used in the original analyses. The uprated design-basis coolant activity provided in Table 11.1-13 is comparable to the original design-basis coolant activities presented in Table 12.2-2. Since the remaining radiation source term data presented in this section are derived from the design-basis coolant concentrations, they remain valid for uprate.

Power uprate has no significant impact on the plant design basis radiation source terms.

12.2.5 References

1. "Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion ALAP for Radioactive Material in Light Water Power Reactor Effluents," WASH-1258, July 1973.

12.2-11 REVISION 15 - DECEMBER 2014

B/B-UFSAR

2. Radiation Analysis Design Manual, WCAP-7664, January 1973.

3. "Source Term Data for Westinghouse Pressurized Water Reactors," WCAP-8253, Pittsburgh, Pa., May 1974.

4. "Oconee Radiochemistry Survey Program," RDTPL75-4, Babcock &

Wilcox, May 1975.

12.2-11a REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE 12.2-1 REACTOR COOLANT NITROGEN-16 ACTIVITY LOCATION IN PRIMARY TIME AFTER LEAVING ACTIVITY COOLANT LOOP THE CORE (sec) (Ci/g) leaving core 0.0 136 leaving reactor vessel 1.3 113 entering steam generator 1.7 109 leaving steam generator 5.8 74 entering reactor coolant pump 6.5 69 entering reactor vessel 7.2 65 entering core 9.2 54 12.2-12

B/B-UFSAR TABLE 12.2-2 REACTOR COOLANT SOURCES FOR SHIELDING DESIGN (ORIGINAL DESIGN BASIS)

ISOTOPE ACTIVITY (Ci/g) ISOTOPE ACTIVITY (Ci/g)

H-3 3.5 (maximum)

Kr-85 8.8 (peak) I-131 2.5 Kr-85m 2.1 I-132 2.8 Kr-87 1.2 I-133 4.0 Kr-88 3.7 I-134 0.6 Xe-131m 1.9 I-135 2.2 Xe-133 281.0 Te-132 0.3 Xe-133m 18.8 Te-134 2.9 x 10-2 Xe-135 6.3 Cs-134 2.3 Xe-135m 0.4 Cs-136 2.8 Xe-138 0.7 Cs-137 1.5 Cs-138 0.98 Br-84 4.3 x 10-2 Ba-137m 1.4 Rb-88 3.7 Ba-140 4.3 x 10-3 Rb-89 0.21 La-140 1.5 x 10-3 Sr-89 3.3 x 10-3 Ce-144 3.4 x 10-4 Sr-90 1.7 x 10-4 Pr-144 3.4 x 10-4 Sr-91 1.9 x 10-3 Sr-92 7.4 x 10-4 Mn-54 7.9 x 10-4 Y-90 2.0 x 10-4 Mn-56 3.0 x 10-2 Y-91 6.1 x 10-3 Co-58 2.6 x 10-2 Y-92 7.2 x 10-4 Co-60 1.0 x 10-3 Zr-95 7.0 x 10-4 Fe-59 1.1 x 10-3 Nb-95 6.9 x 10-4 Cr-51 9.5 x 10-4 Mo-99 5.3 This table is based on the following:

a. Reactor coolant mass = 2.42 x 108 grams.

b. Operation with defects in cladding of rods generating 1% of the core rated power of 3565 MWt.

c. Reactor coolant purification rate = 75 gpm.

d. The average sources expected during normal operation are assumed to be 20% of the maximum values listed.

This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

e. See Table 11.1-13 for the design basis reactor coolant inventory based on the uprated power level.

12.2-13 REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE 12.2-3 DEPOSITED CORROSION PRODUCT ACTIVITY ON STEAM GENERATOR PRIMARY SIDE SURFACES (Ci/cm2)

OPERATING TIME (months)

ISOTOPES 6 12 24 36 Mn-54 0.15 0.60 1.5 2.0 Mn-56 3.3 3.3 3.3 3.3 Co-58 4.5 10.2 11.0 11.0 Fe-59 1.4 3.0 3.0 3.0 Co-60 0.26 1.0 2.6 4.5 12.2-14

B/B-UFSAR TABLE 12.2-4 PRESSURIZER LIQUID PHASE ACTIVITY ISOTOPE ACTIVITY (Ci/g)

N-16 (maximum) 1.3 Rb-88 1.1 x 10-2 Mo-99 2.2 I-131 1.6 I-132 6.2 x 10-2 I-133 7.0 x 10-1 I-134 5.5 x 10-3 I-135 1.4 x 10-1 Cs-134 1.9 Cs-136 2.1 x 10-1 Cs-137 1.3 Cs-138 5.5 x 10-3 Ba-137m 1.2 12.2-15

B/B-UFSAR TABLE 12.2-5 PRESSURIZER STEAM PHASE ACTIVITY ISOTOPE ACTIVITY (Ci/cm3)*

Kr-85 5.1 x l01 Kr-85m 1.0 x l0-1 Kr-87 1.8 x l0-2 Kr-88 1.2 x 10-1 Xe-131m 4.7 Xe-133 3.6 x 102 Xe-133m 10.9 Xe-135 6.5 x 10-1 Xe-135m 1.3 x 10-3 Xe-138 2.2 x 10-3 at operating conditions.

12.2-16

B/B-UFSAR TABLE 12.2-6 PRESSURIZER DEPOSITED ACTIVITY ISOTOPE ACTIVITY (Ci/cm2)

Cr-51 9.8 x 10-2 Mn-54 1.5 x 10-1 Mn-56 2.2 x 10-2 Co-58 3.8 Co-60 2.1 x 10-1 Fe-59 1.4 x 10-1 12.2-17

B/B-UFSAR TABLE 12.2-7 LETDOWN COOLANT ACTIVITY GAMMA ENERGY SPECIFIC SOURCE STRENGTH (MeV/) (MeV/cm3-sec) 0.4 4.5 x 105 0.8 2.7 x 105 1.3 1.7 x 105 1.7 1.2 x 105 2.2 1.4 x 105 2.5 1.6 x 105 3.5 1.9 x 104 NOTE: Same isotopic composition as Table 12.2-2.

12.2-18

B/B-UFSAR TABLE 12.2-8 VOLUME CONTROL TANK VAPOR PHASE (240 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

Kr-85 2.0 1.4 x 101 Kr-85m 9.1 6.2 x 101 Kr-87 3.4 2.3 x 101 Kr-88 1.62 x l01 1.1 x 102 Xe-131m 1.44 x 101 9.8 x 101 Xe-133 2.1 x l03 1.4 x 104 Xe-133m 1.4 x 102 9.5 x 102 Xe-135 3.8 x l01 2.6 x 102 Xe-135m 4.0 x 10-1 2.7 Xe-138 7.0 x 10-1 4.8 LIQUID PHASE (160 ft3)

ISOTOPE ACTIVITY (Ci/g) INVENTORY (Ci)

Kr-85 8.8 (peak) 4.0 x 101 Kr-85m 1.5 6.8 Kr-87 0.5 2.3 Kr-88 2.1 9.5 Xe-131m 1.9 8.6 Xe-133 2.76 x 102 1.2 x l03 Xe-133m 1.8 x 101 7.9 x 101 Xe-135 5.1 2.3 x l01 Rb-88 3.7 x 10-1 1.7 Rb-89 2.1 x 10-2 9.6 x 10-2 Mo-99 5.3 x 10-1 2.4 I-131 2.5 x 10-1 1.1 I-132 2.8 x 10-1 1.3 I-133 4.0 x l0-1 1.8 12.2-19

B/B-UFSAR TABLE 12.2-8 (Cont'd)

LIQUID PHASE (160 ft3)

ISOTOPE ACTIVITY (Ci/g) INVENTORY (Ci)

I-134 5.6 x 10-2 2.5 x 10-1 I-135 2.2 x 10-1 1.0 Cs-134 2.3 x 10-1 1.1 Cs-136 2.8 x 10-1 1.3 Cs-137 1.5 x 10-1 6.8 x 10-1 Cs-138 9.8 x 10-2 4.4 x 10-1 Ba-137m 1.4 x 10-1 6.4 x 10-1 12.2-20

B/B-UFSAR TABLE 12.2-9 RECYCLE HOLDUP TANK VAPOR PHASE SOURCES (8087 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

Kr-85 5.0 1.1 x 103 Kr-85m 0.8 1.8 x 102 Kr-87 0.3 6.9 x 101 Kr-88 1.2 2.7 x 102 Xe-131m 1.0 2.3 x 102 Xe-133 1.4 x 102 3.2 x l04 Xe-133m 9.1 2.1 x 103 Xe-135 2.7 6.2 x 102 Xe-135m 2.4 x 10-2 5.5 Xe-138 4.2 x 10-2 9.6 12.2-21

B/B-UFSAR TABLE 12.2-10 RECYCLE HOLDUP TANK LIQUID PHASE SOURCES (6886 ft3)

ISOTOPE ACTIVITY (Ci/g) INVENTORY (Ci)

H-3 3.5 6.8 x l02 Kr-85 8.8 (peak) 1.7 x 103 Kr-85m 2.1 4.1 x 102 Kr-87 1.2 2.3 x 102 Kr-88 3.7 7.2 x 102 Xe-131m 1.9 3.7 x 102 Xe-133 2.8 x l02 5.5 x 104 Xe-133m 1.9 x l01 3.6 x 103 Xe-135 6.3 1.2 x 103 Rb-88 3.7 x 10-2 7.2 Rb-89 2.1 x l0-3 4.0 x l0-1 Mo-99 5.3 x l0-2 1.0 x 101 I-131 2.5 x 10-2 4.9 I-132 2.8 x 10-2 5.6 I-133 4.0 x 10-2 7.8 I-134 5.6 x 10-3 1.1 I-135 2.2 x 10-2 4.3 Cs-134 2.3 x 10-2 4.4 Cs-136 2.8 x 10-2 5.4 Cs-137 1.5 x 10-2 2.9 Cs-138 9.8 x 10-3 1.9 Ba-137m 1.4 x 10-2 2.7 12.2-22

B/B-UFSAR TABLE 12.2-11 RECYCLE EVAPORATOR VENT CONDENSER SECTION (7900 cm3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

Kr-85 293 2.3 Kr-85m 70 5.5 x 10-1 Kr-87 41 3.2 x 10-1 Kr-88 124 9.8 x 10-1 Xe-131m 64 5.1 x 10-1 Xe-133 9343 7.4 x 10+1 Xe-133m 630 5.0 Xe-135 209 1.7 Xe-135m 13 1.0 x 10-1 Xe-138 23 1.8 x 10-1 EVAPORATOR SECTION (CONCENTRATES) MASS = 2.08 x 106 grams ISOTOPE ACTIVITY (Ci/g) INVENTORY (Ci)

I-131 0.74 1.5 I-132 0.12 2.6 x 10-1 I-133 0.87 1.8 I-134 0.01 2.1 x 10-2 I-135 0.25 5.2 x 10-1 Mo-99 1.48 3.1 Cs-134 0.69 1.5 Cs-137 0.47 9.8 x 10-1 Ba-137m 0.44 9.5 x 10-1 12.2-23

B/B-UFSAR TABLE 12.2-12 RESIDUAL HEAT REMOVAL LOOP RESIDUAL HEAT REMOVAL LOOP SOURCES ISOTOPE ACTIVITY (Ci/g) ISOTOPE ACTIVITY (Ci/g)

Mo-99 4.0 Kr-85 7.6 I-131 1.6 Kr-85m 0.96 I-132 0.56 Kr-87 0.12 I-133 2.3 Kr-88 1.6 x l01 I-135 1.0 Xe-133 2.4 x l02 Te-132 0.17 Xe-133m 2.6 Cs-134 1.8 Xe-135 4.0 Cs-136 2.2 Cs-137 1.2 Ba-137m 1.1 12.2-24

BYRON-UFSAR TABLE 12.2-13 MIXED BED DEMINERALIZER (30 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

Br-84 0.6 5.1 x l0-1 Rb-88 28.9 2.5 x 101 Rb-89 1.3 1.1 Sr-89 1.04 x 102 8.8 x 101 Sr-90 15.2 1.3 x 101 Sr-91 0.5 4.2 x 10-1 Sr-92 5.2 x 10-2 4.4 x 10-2 Y-90 7.5 6.4 Y-91 10.8 9.2 Y-92 0.12 1.0 x 10-1 Zr-95 26.7 2.3 x 101 Nb-95 39.7 3.4 x 101 Mo-99 1.31 x 103 1.1 x 103 I-131 1.25 x 104 1.1 x 104 I-132 1.84 x l03 1.6 x l03 I-133 2.18 x 103 1.9 x 103 I-134 13.4 1.1 x 101 I-135 3.77 x 102 3.2 x 102 Te-132 5.4 x 102 4.6 x 102 Te-134 0.54 4.6 x 10-1 Cs-134 1.01 x l04 8.4 x 103 Cs-136 7.4 x l02 6.2 x 102 Cs-137 6.6 x 103 5.6 x 103 Cs-138 12.9 1.1 x 101 Ba-137m 6.2 x 103 5.2 x 103 Ba-140 34.3 2.9 x 101 La-140 35.8 3.0 x 101 Ce-144 21.8 1.9 x 101 Pr-144 21.8 1.9 x 101 Mn-54 48.0 4.1 x 101 Mn-56 1.2 1.0 Co-58 5065 4300 Co-60 85.0 7.2 x 101 Fe-59 18.0 1.5 x 101 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-25 REVISION 14 - DECEMBER 2012

BRAIDWOOD-UFSAR TABLE 12.2-13 MIXED BED DEMINERALIZER (39 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

Br-84 0.6 6.63 x l0-1 Rb-88 28.9 3.19 x 101 Rb-89 1.3 1.44 Sr-89 1.04 x 102 1.15 x 102 Sr-90 15.2 1.68 x 101 Sr-91 0.5 5.22 x 10-1 Sr-92 5.2 x 10-2 5.74 x 10-2 Y-90 7.5 8.28 Y-91 10.8 1.19 x 101 Y-92 0.12 1.33 x 10-1 Zr-95 26.7 2.95 x 101 Nb-95 39.7 4.38 x 101 Mo-99 1.31 x 103 1.45 x 103 I-131 1.25 x 104 1.38 x 104 I-132 1.84 x l03 2.03 x l03 I-133 2.18 x 103 2.41 x 103 I-134 13.4 1.48 x 101 I-135 3.77 x 102 4.16 x 102 Te-132 5.4 x 102 5.96 x 102 Te-134 0.54 5.96 x 10-1 Cs-134 1.01 x l04 1.12 x 104 Cs-136 7.4 x l02 8.17 x 102 Cs-137 6.6 x 103 7.29 x 103 Cs-138 12.9 1.42 x 101 Ba-137m 6.2 x 103 6.85 x 103 Ba-140 34.3 3.79 x 101 La-140 35.8 3.95 x 101 Ce-144 21.8 2.41 x 101 Pr-144 21.8 2.41 x 101 Mn-54 48.0 5.3 x 101 Mn-56 1.2 1.33 Co-58 5065.0 5590.0 Co-60 85.0 9.39 x 101 Fe-59 18.0 1.99 x 101 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-25a REVISION 14 - DECEMBER 2012

B/B-UFSAR TABLE 12.2-14 CATION BED DEMINERALIZER (20 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

Y-90 11.4 6.5 Y-91 16.2 9.2 Mo-99 1.96 x 103 1.1 x 103 Cs-134 1.5 x 104 8.4 x 103 CS-136 1.1 x 103 6.2 x 102 Cs-137 9.89 x 103 5.6 x 103 Ba-137m 9.2 x 103 5.2 x 103 NOTE:The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-26

B/B-UFSAR TABLE 12.2-15 THERMAL REGENERATION DEMINERALIZER (70 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

I-131 81.0 1.6 x 102 I-132 3.4 6.8 I-133 18.5 3.7 x 101 I-134 0.4 7.9 x 10-1 I-135 4.5 8.9 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-27

B/B-UFSAR TABLE 12.2-16 RECYCLE EVAPORATOR FEED DEMINERALIZER (30 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

I-131 1.22 x 103 1.0 x l03 I-132 1.73 x 101 1.46 x 102 I-133 1.90 x 102 1.6 x l02 I-134 9.6 x 10-1 8.2 x 10-1 I-135 2.98 x 101 2.5 x 101 Cs-134 1.70 x 103 1.5 x 103 Cs-137 1.1 x 103 8.8 x 102 Ba-137m 1.04 x 103 8.3 x 102 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-28

B/B-UFSAR TABLE 12.2-17 RECYCLE EVAPORATOR CONDENSATE DEMINERALIZER (20 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

I-131 3.8 2.2 I-132 5.6 x 10-2 3.1 x 10-2 I-133 0.72 4.1 x 10-1 I-134 4.4 x 10-3 2.5 x 10-3 I-135 0.13 7.4 x 10-2 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-29

B/B-UFSAR TABLE 12.2-18 SPENT FUEL PIT DEMINERALIZER (30 ft3)

ISOTOPE ACTIVITY (Ci/cm3) INVENTORY (Ci)

Co-58 4.2 3.6 I-131 182.4 150.0 Cs-134 14.5 12.3 Cs-137 5.7 4.8 Ba-137m 5.3 4.5 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-30

B/B-UFSAR TABLE 12.2-19 REACTOR COOLANT FILTER ISOTOPE INVENTORY (Ci)

Co-58 8.9 Co-60 2.35 Cs-134 15.0 Cs-137 9.78 Ba-137m 9.16 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-31

B/B-UFSAR TABLE 12.2-20 SEAL WATER RETURN FILTER, RECYCLE EVAPORATOR FEED FILTER, SPENT FUEL PIT FILTER, AND SPENT FUEL PIT SKIMMER FILTER ISOTOPE INVENTORY (Ci)

Co-58 1.78 Co-60 0.47 Cs-134 3.00 Cs-137 1.96 Ba-137m 1.84 SEAL WATER INJECTION FILTER ISOTOPE INVENTORY (Ci)

Co-58 1.17 Co-60 0.30 Cs-134 2.0 Cs-137 1.29 Ba-137m 1.21 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-32

B/B-UFSAR TABLE 12.2-21 RECYCLE EVAPORATOR CONCENTRATE FILTER ISOTOPE INVENTORY (Ci)

Co-58 2.2 x 10-2 Co-60 5.9 x 10-3 Cs-134 3.7 x 10-2 Cs-137 2.4 x 10-2 Ba-137m 2.2 x 10-2 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-33

B/B-UFSAR TABLE 12.2-22 RECYCLE EVAPORATOR CONDENSATE FILTER ISOTOPE INVENTORY (Ci)

I-131 2.15 x 10-2 I-132 3.1 x 10-4 I-133 4.1 x 10-3 I-134 2.5 x l0-5 I-135 7.1 x l0-4 NOTE: The average sources expected during normal operation are assumed to be 20% of the maximum values listed. This average is based on operating experience to date with Westinghouse's PWRs using Zircaloy-clad fuel.

12.2-34

B/B-UFSAR TABLE 12.2-23 CORE SHUTDOWN SOURCES - (MeV/cm3-sec)

TIME AFTER SHUTDOWN PHOTON ENERGY (MeV) 4 HOURS 12 HOURS 1 DAY 1 WEEK 1 MONTH 3 MONTHS 0.4 3.1 X 1011 2.3 x 1011 1.9 x 1011 9.2 x 1010 3.8 x 1010 1.3 x 1010 0.8 1.3 X 1012 9.8 x 1011 8.0 x 1011 4.0 x 1011 2.3 x 1011 1.2 x 1011 1.3 3.9 X 1011 2.9 x 1011 2.5 x 1011 1.6 x 1011 1.2 x 1011 5.8 x 1010 1.7 5.1 X 1011 3.8 x 1011 3.3 x 1011 2.3 x 1011 6.2 x 1010 2.9 x 109 2.2 7.2 X 1010 2.6 x 1010 1.5 x 1010 8.5 x 109 6.7 x 109 5.0 x 109 2.5 8.9 X 1010 4.7 x 1010 3.7 x 1010 2.5 x 1010 7.9 x 109 3.5 x 108 3.5 8.2 X 109 2.0 x 109 1.3 x 109 9.6 x 108 2.0 x 108 1.5 x 107 12.2-35

B/B-UFSAR TABLE 12.2-24 IRRADIATED Ag-In-Cd CONTROL ROD SOURCES (Ci/cm/rod)

TIME AFTER SHUTDOWN ISOTOPE 0 1 WEEK 1 MONTH 6 MONTHS 1 YEAR Ag-110m 50 49 46 30 18 12.2-36

B/B-UFSAR TABLE 12.2-24a HAFNIUM CONTROL ROD SOURCE STRENGTHS 400-DAY IRRADIATION SOURCE STRENGTH AT TIME AFTER SHUTDOWN (Mev/cm-s)

ENERGY GROUP (MeV) 1 DAY 1 WEEK 1 MONTH 6 MONTHS 1 YEAR 0.20 - 0.40 2.2 x 1010 2.0 x 1010 1.4 x 1010 1.3 x 109 1.0 x 108 0.40 - 0.90 1.9 x 1011 1.7 x 1011 1.2 x 1011 1.0 x 1010 5.0 x 108 0.90 - 1.35 2.6 x 1010 2.5 x 1010 2.2 x 1010 8.9 x 109 2.9 x 109 15-YEAR IRRADIATION SOURCE STRENGTH AT TIME AFTER SHUTDOWN (Mev/cm-s)

ENERGY GROUP (MeV) 1 DAY 1 WEEK 1 MONTH 6 MONTHS 1 YEAR 0.20 - 0.40 4.7 x 1010 4.3 x 1010 2.9 x 1010 3.8 x 109 6.2 x 108 0.40 - 0.90 3.5 x 1011 3.2 x 1011 2.2 x 1011 1.9 x 1010 9.3 x 108 0.90 - 1.35 2.8 x 1011 2.7 x 1011 2.4 x 1011 9.7 x 1010 3.1 x 1010

Source strengths are expressed per cm3 of absorber. Density of the hafnium absorber is 13.31 g/cm3.

12.2-37

B/B-UFSAR TABLE 12.2-25 REFUELING WATER ACTIVITY CONCENTRATIONS RESULTING IN 2.5 mrem/hr AT THE SURFACE MAXIMUM ALLOWABLE DOMINANT CONCENTRATION SOURCE OF ACTIVITY ISOTOPE (CI/cm3)

A. Fission Product Gases Xe-133 0.15 B. Fission Product Particulates Cs-137 0.005 C. Corrosion Products Co-58 0.005 D. Fission Product Halogens I-131 0.01 12.2-38

B/B-UFSAR TABLE 12.2-26 INCORE INSTRUMENTS - FISSION CHAMBER SOURCES*

(Bases: Irradiation Period = 3 Months Decay Period = 1 Day)

GAMMA ENERGY GROUP ACTIVITY MeV (MeV/sec) 0.4 8.1 x 109 0.8 4.7 X 1010 1.3 7.5 X 108 1.7 1.9 X 1010 2.2 5.0 X 108 2.5 1.7 X 109 3.5 5.0 X 107 Spectrum represents irradiated Ag-110m.

12.2-39

B/B-UFSAR TABLE 12.2-27 DRIVE WIRE SOURCES (Bases: Irradiation Period = 1 Year No Decay)

ACTIVITY ISOTOPE (Ci/cm)

Mn-54 2.78 x 104 Mn-56 6.48 x 105 Fe-59 2.04 x 104 Co-58 5.41 x 103 Co-60 3.08 x 103 12.2-40

B/B-UFSAR TABLE 12.2-28 SINGLE WASTE GAS DECAY TANK ACTIVITIES (Single Unit)

ISOTOPE ACTIVITY (Ci)

Kr-85 6.3 x 102 (peak)

Kr-85m 1.3 x 102 Kr-87 2.0 x 101 Kr-88 1.7 x 102 Xe-131m 2.2 x 102 Xe-133 3.2 x 104 Xe-133m 2.2 x 103 Xe-135 5.4 x 102 Xe-135m less than 1 Xe-138 less than 1 12.2-41

B/B-UFSAR TABLE 12.2-29 SPENT FUEL PIT WATER ACTIVITY FOR A FUEL HANDLING ACCIDENT ISOTOPE ACTIVITY (Ci/cm3)

I-131 13.1 I-132 7.16 I-133 10.4 I-134 2.34 I-135 5.31 Kr-85m 1.65 Kr-85 1.77 Kr-87 0.969 Kr-88 4.47 Xe-131m 0.120 Xe-133 23.9 Xe-33m 2.51 Xe-135 1.88 Xe-135m 0.177 Xe-138 1.06 NOTE: These activities are the maximum gap activities of one fuel assembly distributed in the fuel pit water.

12.2-42

B/B-UFSAR TABLE 12.2-30 SHIELDING DESIGN-BASIS INFLUENT RADIOACTIVITY CONCENTRATION IN LIQUID WASTE PROCESSING STREAMS STEAM GENERATOR CHEMICAL REGEN. WASTE BLOWDOWN** DRAINS DRAINS LAUNDRY DRAINS ISOTOPE (Ci/gm) (Ci/gm) (Ci/gm) (Ci/gm)

H-3 2.6-02* 3.5-03 3.5-02 3.5-06 Na-24 2.4-07 2.0-03 - 2.9-06 Cr-51 7.1-06 9.6-07 9.6-06 9.6-10 Mn-54 5.9-06 7.9-07 7.9-06 3.5-05 Mn-56 2.2-04 3.0-05 3.0-04 3.0-08 Co-58 1.9-04 2.6-05 2.6-04 7.3-05 Fe-59 8.5-06 1.1-06 1.1-05 1.5-06 Co-60 7.0-02 1.0-06 1.0-05 2.6-05 Br-84 3.2-04 4.3-05 4.3-04 4.3-08 Rb-88 2.8-02 3.7-03 3.7-02 3.7-06 Rb-89 1.6-03 2.1-04 2.1-03 2.1-07 Sr-89 2.5-05 3.3-06 3.3-05 3.3-09 Sr-90 1.3-06 1.7-07 1.7-06 2.0-06 Y-90 1.5-06 2.0-07 2.0-06 2.0-10 Sr-91 1.4-05 1.9-06 1.9-05 1.9-09 Y-91 4.5-05 6.1-06 6.1-05 6.1-09 Sr-92 5.5-06 7.4-07 7.4-06 7.4-10 Y-92 4.5-06 7.2-07 7.2-06 7.2-10 Y-93 - 8.0-05 - -

Zr-95 5.2-06 7.0-07 7.0-06 7.0-10 Nb-95 5.1-06 6.9-07 6.9-06 6.9-10 Mo-99 4.0-02 5.3-03 5.3-02 5.3-06 Sb-124 - 1.2-05 - -

I-131 1.9-02 2.5-03 2.5-02 5.0-04 Te-132 1.6-03 2.3-04 2.3-03 2.3-07 I-132 2.1-02 2.8-03 2.8-02 2.8-06 I-133 3.0-02 4.0-03 4.0-02 4.0-06 Te-134 2.1-04 2.9-05 2.9-04 2.9-08 I-134 4.4-03 6.0-04 6.0-03 6.0-07 Cs-134 1.7-02 2.3-03 2.3-02 1.0-05 I-135 1.6-02 2.2-03 2.2-02 2.2-06 Cs-136 2.1-02 2.8-03 2.8-02 2.8-06 Cs-137 1.1-02 1.5-03 1.5-02 2.0-05 Ba-137m 1.0-02 1.4-03 1.4-02 1.9-05 Cs-138 7.1-03 9.8-04 9.8-03 9.8-07 Ba-140 3.2-05 4.3-06 4.3-05 4.3-09 La-140 1.1-05 1.5-06 1.5-05 1.5-09 Ce-141 - 1.7-07 - -

Ce-144 2.5-06 3.4-07 3.4-06 3.4-10 Pr-144 2.5-06 3.4-07 3.4-06 3.4-10

2.6-02 means 2.6 x 10-2

- Shielding was determined based on equipment 1/2WX02MA,B (housing-only prefilter vessels).

12.2-43 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.2-31 SHIELDING DESIGN-BASIS INFLUENT RADIOACTIVITY CONCENTRATIONS IN LIQUID WASTE PROCESSING STREAMS TURBINE AUX. BLDG. BUILDING EQUIPMENT AUX. BLDG. EQUIPMENT TURBINE BLDG.

DRAINS FLOOR DRAINS DRAINS FLOOR DRAINS ISOTOPE (Ci/gm) (Ci/gm) (Ci/gm) (Ci/gm)

H-3 7.0-01* 7.0-02 1.3-04 1.3-04 C-14 4.0-06 8.0-06 - -

Na-24 3.0-04 2.0-05 - -

Cr-51 1.9-04 1.9-05 3.6-09 3.6-09 Mn-54 1.6-04 1.6-05 2.9-09 2.9-09 Fe-55 2.0-04 1.0-03 - -

Mn-56 6.0-03 6.0-04 1.1-07 1.1-07 Co-58 5.2-03 5.2-04 9.6-08 9.6-08 Fe-59 2.2-04 2.2-05 1.5-09 1.5-09 Co-60 2.0-04 2.0-05 3.6-09 3.6-09 Ni-63 4.0-05 8.0-05 - -

Br-84 8.6-03 8.6-04 1.6-06 1.6-06 Rb-88 7.4-01 7.4-02 1.4-05 1.4-05 Rb-89 4.2-02 4.2-03 7.8-07 7.8-07 Sr-89 6.6-04 6.6-05 1.2-08 1.2-08 Sr-90 3.4-05 3.4-06 6.3-10 6.3-10 Y-90 4.0-05 4.0-06 7.4-10 7.4-10 Sr-91 3.8-04 3.8-05 7.0-09 7.0-09 Y-91 1.2-03 1.2-04 2.3-08 2.3-08 Sr-92 1.5-04 1.5-05 2.7-09 2.7-09 Y-92 1.4-04 1.4-05 2.7-09 2.7-09 Zr-95 1.4-04 1.4-05 2.6-09 2.6-09 Nb-95 1.4-04 1.4-05 2.6-09 2.6-09 Mo-99 1.1-00 1.1-01 2.0-05 2.0-05 Ru-103 2.0-05 1.0-06 - -

Sb-124 6.0-05 1.6-05 - -

I-131 5.0-01 5.0-02 9.3-05 9.3-05 Te-132 4.5-02 4.5-03 8.3-07 8.3-07 I-132 5.6-01 5.6-02 1.0-04 1.0-04 I-133 8.0-01 8.0-02 1.5-04 1.5-04 Te-134 5.8-03 5.8-04 1.1-07 1.1-07 I-134 1.2-01 1.2-02 2.2-05 2.2-05 7.0-01 means 7.0 x 10-1 12.2-44

B/B-UFSAR TABLE 12.2-31 (Cont'd)

TURBINE AUX. BLDG. BUILDING EQUIPMENT AUX. BLDG. EQUIPMENT TURBINE BLDG.

DRAINS FLOOR DRAINS DRAINS FLOOR DRAINS ISOTOPE (Ci/gm) (Ci/gm) (Ci/gm) (Ci/gm)

Cs-134 4.6-01 4.6-02 8.4-06 8.4-06 I-135 4.4-01 4.4-02 8.1-05 8.1-05 Cs-136 5.6-01 5.6-02 1.0-05 1.0-05 Cs-137 3.0-01 3.0-02 5.6-06 5.6-06 Ba-137m 2.8-01 2.8-02 5.2-06 5.2-06 Cs-138 2.0-01 2.0-02 3.6-06 3.6-06 Ba-140 8.6-04 8.6-05 1.6-08 1.6-08 La-140 3.0-04 3.0-05 5.6-09 5.6-09 Ce-144 6.8-05 6.8-06 1.3-09 1.3-09 Pr-144 6.8-05 6.8-06 1.3-09 1.3-09 12.2-45

B/B-UFSAR TABLE 12.2-32 SOURCE BASES FOR DRAIN TANKS TOTAL MAXIMUM FRACTION OF TANK DAILY FLOW PRIMARY COOLANT NAME QUANTITY (gal/day) CONTAINED Turbine Bldg.

Floor Drain Tank 2 12,000 3.704 x 10-5 for Iodines 3.704 x 10-6 for non-Iodines Turbine Bldg.

Equipment Drains Tank 2 12,000 3.704 x 10-5 for Iodines 3.704 x 10-6 for non-Iodines Aux. Bldg.

Floor Drain Tank 2 16,000 .02 Aux. Bldg.

Equipment Drain Tank 2 16,000 .2 Chemical Drain Tank 1 6,000 .001 Chemical/Regeneration Waste Drain Tank 1 10,000 .01 Laundry Drain Tank 1 4,000 1 x 10-6

+ Table 12.2-33 Sources.

12.2-46 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.2-33 LAUNDRY DRAIN SOURCES USED IN SHIELDING SOURCE CALCULATION ISOTOPE ISOTOPIC ACTIVITY (Ci/cc)

Na-24 2.9 x 10-6 Mn-54 3.5 x 10-5 Co-58 7.3 x 10-5 Fe-59 1.5 x 10-6 Co-60 2.6 x 10-5 Sr-90 2.0 x 10-6 I-131 5.0 x 10-4 Cs-134 1.0 x 10-5 Cs-137 2.0 x 10-5 Ba-137m 1.9 x 10-5 NOTE: In addition, 1 x 10-6 x primary coolant activity is added to the above inventory.

12.2-47 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-34 DECONTAMINATION FACTORS USED IN SHIELDING SOURCE CALCULATION OF LIQUID RADWASTE PROCESSING SYSTEM AND BLOWDOWN SYSTEM COMPONENTS ATOMIC COMPONENT NUMBER ELEMENT FILTER DEMINERALIZER EVAPORATOR 1 H 1 1 1 6 C 1 1 10000 11 Na 1 1 10000 24 Cr 10 10 10000 25 Mn 10 10 10000 26 Fe 10 10 10000 27 Co 10 10 10000 28 Ni 10 10 10000 35 Br 1 10 10000 36 Kr 1 1 1 37 Rb 1 1 10000 38 Sr 1 10 10000 39 Y 1 1 10000 40 Zr 10 10 10000 41 Nb 10 10 10000 42 Mo 1 1 10000 44 Ru 10 10 10000 51 Sb 1 1 10000 52 Te 1 10 10000 53 I 1 10 1000 54 Xe 1 1 1 55 Cs 1 1 10000 56 Ba 1 10 10000 57 La 1 10 10000 58 Ce 10 10 10000 59 Pr 10 10 10000 12.2-48

B/B-UFSAR TABLE 12.2-35 SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT IN LIQUID RADWASTE PROCESSING SYSTEM COMPONENTS (Curies)

BLOWDOWN MIXED RADWASTE MIXED ISOTOPE BED DEMINERALIZER BED DEMINERALIZER Cr-51 5.3 x 10-3 2.3 x 10-6 Mn-54 5.1 x 10-3 2.6 x 10-6 Mn-56 2.2 x 10-3 -

Co-58 1.6 x 10-1 7.5 x 10-5 Fe-59 6.5 x 10-3 2.9 x 10-6 Co-60 5.0 x 10-3 2.6 x 10-6 Br-84 7.0 x 10-3 2.1 x 10-6 Rb-88 4.8 x 10-2 5.1 x 10-5 Rb-89 1.4 x 10-3 1.5 x 10-6 Sr-89 2.0 x 10-1 9.1 x 10-5 Sr-90 1.1 x 10-2 5.7 x 10-6 Y-90 8.2 x 10-3 4.9 x 10-6 Sr-91 5.2 x 10-3 1.3 x 10-6 Sr-92 5.8 x 10-4 -

Y-92 5.8 x 10-4 -

Zr-95 4.3 x 10-3 2.0 x 10-6 Nb-95 4.5 x 10-3 2.3 x 10-6 Mo-99 6.9 x 10-2 7.4 x 10-5 I-131 9.6 x 10-1 3.0 x 10-1 Xe-131m 2.1 x 10-1 1.2 x 10-3 Te-132 5.4 1.4 x 10-3 I-132 6.0 2.9 x 10-3 I-133 2.3 x 101 5.7 x 10-2 Xe-133m 5.6 x 10-1 1.4 x 10-3 Xe-133 1.9 x 101 5.5 x 10-2 Te-134 6.1 x 10-3 1.8 x 10-6 I-134 1.6 x 10-1 4.3 x 10-4 Cs-134 3.9 x 10-3 4.2 x 10-6 BASES:

1. Time period of collection is 14 days for the blowdown demineralizer and 30 days for the radwaste demineralizer.

2. 1% failed fuel.

12.2-49

B/B-UFSAR TABLE 12.2-35 (Cont'd)

BLOWDOWN MIXED RADWASTE MIXED ISOTOPE BED DEMINERALIZER BED DEMINERALIZER I-135 4.2 1.0 x 10-2 Xe-135m 1.2 3.0 x 10-3 Xe-135 4.2 9.9 x 10-3 Cs-136 2.0 x 10-3 2.1 x 10-6 Cs-137 2.0 x 10-2 2.1 x 10-5 Cs-138 1.3 x 10-2 1.4 x 10-5 Ba-140 2.0 x 10-1 7.1 x 10-5 La-140 1.9 x 10-1 7.2 x 10-5 Ce-144 2.2 x 10-3 1.1 x 10-6 Pr-144 2.2 x 10-3 1.1 x 10-6 BASES:

1. Time period of collection is 14 days for the blowdown demineralizer and 30 days for the radwaste demineralizer.

2. 1% failed fuel.

12.2-50

BYRON-UFSAR TABLE 12.2-36 SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT IN LIQUID RADWASTE PROCESSING SYSTEM COMPONENTS (Curies)

ISOTOPE CONCENTRATES HOLDING TANK C-14 1.9 x 10-2 Na-24 5.1 x 10-2 Cr-51 4.8 x 10-3 Mn-54 4.5 x 10-3 Fe-55 1.9 Mn-56 5.4 x 10-3 Co-58 1.4 x 10-1 Fe-59 5.7 x 10-3 Co-60 5.6 x 10-3 Ni-63 1.9 Br-84 1.6 x 10-2 Rb-88 7.8 x 10-1 Rb-89 3.8 x 10-2 Sr-89 1.7 x 10-1 Sr-90 9.6 x 10-3 Y-90 1.0 x 10-2 Sr-91 1.2 x 10-2 Y-91m 6.8 x 10-3 Y-91 3.2 x 10-1 Sr-92 1.4 x 10-3 Y-92 3.1 x 10-3 Y-93 9.6 x 10-4 Zr-95 3.7 x 10-3 Nb-95 3.8 x 10-3 Mo-99 1.4 x 102 Tc-99m 1.1 x 102 Ru-103 3.7 x 10-3 Sb-124 1.1 x 10-1 I-131 9.9 x 101 Xe-131m 1.3 x 10-1 Te-132 6.3 I-132 2.6 x 101 I-133 4.7 x 101 Xe-133m 4.1 Xe-133 1.8 x 101 Te-134 1.4 x 10-2 I-134 3.8 x 10-1 BASES

1. Time collection period is 30 days.

2. 1% failed fuel.

12.2-51 REVISION 1 - DECEMBER 1989

BYRON-UFSAR TABLE 12.2-36 (Cont'd)

ISOTOPE CONCENTRATES HOLDING TANK Cs-134 1.3 x 102 I-135 9.7 Xe-135m 1.7 Xe-135 8.7 Cs-136 1.3 x 102 Cs-137 8.4 x 101 Ba-137m 7.8 x 101 Cs-138 3.7 x 10-1 Ba-140 1.9 x 10-1 La-140 1.6 x 10-1 Ce-141 7.4 x 10-6 Ce-144 1.9 x 10-3 Pr-144 1.9 x 10-3 BASES

1. Time collection period is 30 days.

2. 1% failed fuel.

12.2-52 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-37 SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT IN RADWASTE FILTERS (in Curies)

TURBINE BUILDING TURBINE BUILDING BLOWDOWN BLOWDOWN RADWASTE EQUIPMENT DRAIN FLOOR DRAIN ISOTOPE PREFILTER* AFTERFILTER AFTERFILTER FILTER FILTER Cr-51 7.4 x 10-2 5.3 x 10-4 2.3 x 10-7 3.1 x 10-6 3.1 x 10-6 Mn-54 7.1 x 10-2 5.1 x 10-4 2.6 x 10-7 3.5 x 10-6 3.5 x 10-6 Fe-55 - - 1.8 x 10-5 - -

Mn-56 3.0 x 10-2 2.2 x 10-4 - 7.0 x 10-7 7.0 x 10-7 Co-58 2.2 1.6 x 10-2 7.5 x 10-6 1.0 x 10-4 1.0 x 10-4 Fe-59 9.1 x 10-2 6.5 x 10-4 2.9 x 10-7 4.0 x 10-6 4.0 x 10-6 Co-60 9.1 x 10-2 6.5 x 10-4 3.4 x 10-7 4.6 x 10-6 4.6 x 10-6 Ni-63 - - 1.8 x 10-6 - -

Zr-95 6.0 x 10-2 4.3 x 10-4 2.7 x 10-7 2.7 x 10-6 2.7 x 10-6 Nb-95 6.3 x 10-2 4.5 x 10-4 2.2 x 10-7 3.1 x 10-6 3.1 x 10-6 Ru-103 - - 3.4 x 10-8 - -

Sb-124 - - 6.9 x 10-11 - -

I-131 1.6 x 10-4 1.5 x 10-5 1.7 x 10-7 8.1 x 10-7 8.1 x 10-7 Ce-144 3.1 x 10-2 2.2 x 10-4 1.1 x 10-7 1.5 x 10-6 1.5 x 10-6 Pr-144 3.1 X 10-2 2.2 X 10-4 1.1 X 10-7 1.5 X 10-6 1.5 X 10-6 BASES:

1. Maximum daily flow rates given in Table 12.2-32 except for the blowdown stream which has maximum flow rate of 135 gpm.

2. 1% failed fuel.

3. Time period of collection is 14 days for the blowdown filters and 30 days for the radwaste and turbine building filters.

4. Primary to secondary steam generator leakage of 1 gpm.

Shielding was determined based on equipment 1/2WX02MA,B (housing-only prefilter vessels).

12.2-53 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.2-38 SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT IN RADWASTE FILTERS (in Curies)

AUX. BLDG. AUX. BLDG. REGENERATION EQUIP. DRAIN FLOOR DRAIN WASTE DRAIN CHEMICAL LAUNDRY ISOTOPE FILTER FILTER FILTER DRAIN FILTER DRAIN FILTER Cr-51 2.2 x 10-1 2.2 x 10-2 6.9 x 10-3 4.2 x 10-4 -

Mn-54 2.5 x 10-1 2.5 x 10-2 7.8 x 10-3 4.7 x 10-4 7.9 x 10-3 Fe-55 3.3 x 10-1 1.6 - - -

Mn-56 5.1 x 10-2 5.1 x 10-3 1.6 x 10-3 9.5 x 10-5 -

Co-58 7.4 7.4 x 10-1 2.3 x 10-1 1.4 x 10-2 2.6 x 10-2 Fe-59 2.9 x 10-1 2.9 x 10-2 9.0 x 10-3 5.4 x 10-4 -

Co-60 3.3 x 10-1 3.3 x 10-2 1.0 x 10-3 6.1 x 10-4 4.1 x 10-3 Ni-63 6.5 x 10-2 1.3 x 10-1 - - -

Zr-95 2.0 x 10-6 2.0 x 10-7 6.1 x 10-3 3.7 x 10-4 -

Nb-95 2.2 x 10-1 2.2 x 10-2 6.9 x 10-3 4.1 x 10-4 -

Mo-99 9.3 x 10-3 9.3 x 10-4 4.6 x 10-4 4.6 x 10-5 -

Ru-103 2.5 x 10-3 1.3 x 10-3 - - -

Sb-124 - - - 2.9 x 10-11 -

I-131 4.4 x 10-3 4.4 x 10-4 2.2 x 10-4 2.2 x 10-5 -

I-132 5.0 x 10-3 5.0 x 10-4 2.4 x 10-4 2.4 x 10-5 -

I-133 7.0 X 10-3 7.0 X 10-4 3.5 X 10-4 3.5 X 10-5 -

I-134 1.1 X 10-3 1.1 X 10-4 5.3 X 10-5 5.3 X 10-6 -

I-135 3.9 X 10-3 3.9 X 10-4 1.9 X 10-4 1.9 X 10-5 -

Cs-137 2.6 x 10-3 2.6 x 10-4 1.3 x 10-4 1.3 x 10-5 -

Cs-138 1.7 x 10-3 1.7 x 10-4 - - -

Ce-144 1.1 x 10-1 1.1 x 10-2 3.4 x 10-3 2.0 x 10-4 -

Pr-144 1.1 x 10-1 1.1 x 10-2 3.4 x 10-3 2.0 x 10-4 -

BASES:

1. Maximum daily flow rates given in Table 12.2-32.

2. 1% failed fuel.

3. Time period of collection is 30 days.

4. Primary to secondary steam generator leakage of 1 gpm.

12.2-54

B/B-UFSAR TABLE 12.2-39 SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT IN THE LIQUID RADWASTE PROCESSING SYSTEM COMPONENTS (Curies)

ISOTOPE RADWASTE EVAPORATOR C-14 1.96 x 10-2 Na-24 5.30 x 10-2 Cr-51 4.97 x 10-3 Mn-54 4.52 x 10-3 Fe-55 1.97 Mn-56 5.56 x 10-3 Co-58 1.44 x 10-1 Fe-59 5.93 x 10-3 Co-60 5.77 x 10-3 Ni-63 1.96 x 10-1 Br-84 1.68 x 10-2 Rb-88 8.12 x 10-1 Rb-89 3.99 x 10-2 Sv-89 1.81 x 10-1 Y-89m 1.81 x 10-5 Sr-90 9.82 x 10-3 Y-90 1.06 x 10-2 Sr-91 1.21 x 10-2 Y-91m 7.05 x 10-3 Y-91 3.35 x 10-1 Sr-92 1.44 x 10-3 Y-92 3.18 x 10-3 Y-93 1.0 x 10-3 Zr-95 3.86 x 10-3 Nb-95m 3.57 x 10-5 Nb-95 3.97 x 10-3 Mo-99 1.42 x 10-2 Tc-99m 1.15 x 10-2 Tc-99 5.58 x 10-6 Ru-103 3.8 x 10-3 Sb-124 1.11 x 10-1 I-131 1.03 x 10-2 Xe-131m 1.33 x 10-1 Te-132 6.53 I-132 2.73 x 101 I-133 4.91 x 101 Xe-133m 4.21 Xe-133 1.85 x 101 Te-134 1.49 x 10-2 I-134 3.97 x 10-1 Cs-134 1.33 x 102 I-135 1.01 x 101 Xe-135m 1.72 Xe-135 9.01 12.2-55

B/B-UFSAR TABLE 12.2-39 (Cont'd)

ISOTOPE RADWASTE EVAPORATOR Cs-135 2.66 x 10-8 Cs-136 1.30 x 102 Cs-137 8.67 x 101 Ba-137m 8.10 x 101 Cs-138 3.88 x 10-1 Ba-140 1.98 x 10-1 La-140 1.60 x 10-1 Ce-141 7.70 x 10-6 Ce-144 1.94 x 10-3 Pr-144 1.94 x 10-3 12.2-56

B/B-UFSAR TABLE 12.2-40 SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT IN LIQUID RADWASTE PROCESSING SYSTEM COMPONENTS (Curies) 30,000-GALLON LAUNDRY ISOTOPE RELEASE TANK DRAIN TANK H-3 3.51 x 10-3 2.65 x 10-5 Na-24 2.59 x 10-4 2.19 x 10-5 Cr-51 1.73 x 10-8 7.26 x 10-10 Mn-54 3.10 x 10-4 2.61 x 10-5 Mr-56 5.41 x 10-7 2.27 x 10-8 Co-58 6.47 x 10-4 5.51 x 10-5 Fe-59 1.33 x 10-5 1.10 x 10-6 Co-60 2.31 x 10-4 1.97 x 10-5 Pr-84 4.32 x 10-5 3.25 x 10-7 Rb-88 6.67 x 10-4 2.80 x 10-5 Rb-89 3.78 x 10-5 1.59 x 10-6 Sr-89 5.95 x 10-7 2.50 x 10-8 Sr-90 1.78 x 10-4 1.51 x 10-5 Y-90 3.61 x 10-8 1.51 x 10-9 Sr-91 3.43 x 10-7 1.44 x 10-8 Y-91 1.10 x 10-6 4.62 x 10-8 Sr-92 1.33 x 10-7 5.60 x 10-9 Y-92 1.30 x 10-7 5.45 x 10-9 Zr-95 1.26 x 10-8 5.30 x 10-10 Nb-95 1.24 x 10-8 5.22 x 10-10 Mo-99 9.56 x 10-4 4.01 x 10-5 I-131 6.28 x 10-3 4.00 x 10-3 Te-132 4.07 x 10-5 1.71 x 10-6 I-132 2.55 x 10-4 2.10 x 10-5 I-133 4.01 x 10-3 3.03 x 10-5 Te-134 5.23 x 10-6 2.20 x 10-7 I-134 6.02 x 10-4 4.54 x 10-6 Cs-134 9.16 x 10-4 7.57 x 10-5 I-135 2.21 x 10-3 1.66 x 10-5 Cs-136 5.04 x 10-4 2.12 x 10-5 Cs-137 1.91 x 10-3 1.51 x 10-4 Ba-137m 1.78 x 10-3 1.41 x 10-4 Cs-138 1.77 x 10-4 7.42 x 10-6 Ba-140 7.75 x 10-7 3.25 x 10-8 La-140 2.70 x 10-7 1.14 x 10-8 Ce-144 6.13 x 10-9 2.57 x 10-10 Pr-144 6.13 x 10-9 2.57 x 10-10 12.2-57

B/B-UFSAR TABLE 12.2-41 SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT IN LIQUID RADWASTE PROCESSING SYSTEM COMPONENTS (Curies)

BLOWDOWN RADWASTE EVAPORATOR ISOTOPE MONITOR TANK MONITOR TANK H-3 1.88 2.00 x 101 C-14 - 9.10 x 10-8 Na-24 1.82 x 10-5 1.74 x 10-5 Cr-51 5.14 x 10-7 5.49 x 10-10 Mn-54 4.23 x 10-7 4.52 x 10-10 Fe-55 - 5.5 x 10-6 Mn-56 1.61 x 10-5 1.71 x 10-8 Co-58 1.39 x 10-5 1.49 x 10-8 Fe-59 5.90 x 10-7 6.29 x 10-10 Co-60 5.37 x 10-7 5.72 x 10-10 Ni-63 - 9.08 x 10-9 Br-84 2.31 x 10-3 2.46 x 10-6 Rb-88 1.98 2.12 x 10-3 Rb-89 1.13 x 10-1 1.20 x 10-4 Sr-89 1.77 x 10-4 1.89 x 10-7 Sr-90 9.11 x 10-6 9.72 x 10-9 Y-90 1.07 x 10-4 1.14 x 10-7 Sr-91 1.02 x 10-4 1.09 x 10-7 Y-91 3.27 x 10-3 3.49 x 10-6 Sr-92 3.97 x 10-5 4.23 x 10-8 Y-92 3.86 x 10-4 4.12 x 10-7 Y-93 - 6.07 x 10-7 Zr-95 3.75 x 10-7 4.00 x 10-10 Nb-95 3.70 x 10-7 3.94 x 10-10 Mo-99 2.84 3.03 x 10-3 Ru-103 - 2.25 x 10-10 Sb-124 - 6.60 x 10-7 I-131 1.34 x 10-1 1.43 x 10-3 Te-132 1.21 x 10-2 1.29 x 10-5 I-132 1.49 x 10-1 1.59 x 10-3 I-133 2.14 x 10-1 2.29 x 10-3 Te-134 1.55 x 10-3 1.66 x 10-6 I-134 3.22 x 10-2 3.43 x 10-4 Cs-134 1.23 1.31 x 10-3 I-135 1.18 x 10-1 1.26 x 10-3 Cs-136 1.50 1.60 x 10-3 Cs-137 8.04 x 10-2 8.57 x 10-4 Ba-137m 7.48 x 10-2 7.97 x 10-3 Cs-138 5.25 x 10-1 5.60 x 10-4 Ba-140 2.30 x 10-4 2.46 x 10-7 La-140 8.04 x 10-5 8.57 x 10-8 Ce-141 - 2.9 x 10-11 Ce-144 1.82 x 10-7 1.94 x 10-10 Pr-144 1.82 x 10-7 1.94 x 10-10 12.2-58

BYRON-UFSAR TABLE 12.2-42 ASSUMED DEMINERALIZER RESIN INVENTORY IN SPENT RESIN TANK FOR SHIELDING SOURCES CALCULATION VOLUME OF FRACTIONAL COMPONENT RESIN IN CONTRIBUTION TO QUANTITY NAME EACH (ft3) SRST INVENTORY 4 Letdown Mixed Bed Demineralizers 30 .16 2 Cation Demineralizers 20 .053 2 Recycle Evaporator Feed Demineralizer 30 .08 2 Recycle Evaporator Condensate Demineralizer 20 .053 7 Boron Thermal Regeneration Demineralizers 70 .65 12.2-59 REVISION 1 - DECEMBER 1989

BRAIDWOOD-UFSAR TABLE 12.2-42 ASSUMED DEMINERALIZER RESIN INVENTORY IN HIGH ACTIVITY SPENT RESIN TANK FOR SHIELDING SOURCES CALCULATION VOLUME OF FRACTIONAL COMPONENT RESIN IN CONTRIBUTION TO QUANTITY NAME EACH (ft3) SRST INVENTORY 4 Letdown Mixed Bed Demineralizers 35 .182 2 Cation Demineralizers 20 .052 2 Recycle Evaporator Feed Demineralizer 30 .078 2 Recycle Evaporator Condensate Demineralizer 20 .052 7 Boron Thermal Regeneration Demineralizers 70 .636 12.2-59a REVISION 14 - DECEMBER 2012

BYRON-UFSAR TABLE 12.2-43 SPENT RESIN TANK SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT (in Curies)

ACTIVITY (Curies) ACTIVITY (Curies)*

ISOTOPE NO DECAY 90-DAY DECAY Br-84 2.0 -

Rb-88 9.8 x 101 -

Rb-89 4.55 -

Sr-89 3.54 x 102 1.03 x 102 Sr-90 5.2 x 101 5.2 x 101 Sr-91 1.7 -

Sr-92 1.8 x 10-1 -

Y-90 3.84 x 101 5.1 x 101 Y-91 5.51 x 101 1.9 x 101 Y-92 4.1 x 10-1 -

Zr-95 9.1 x 101 3.5 x 101 Nb-95 1.3 x 102 6.5 x 101 Mo-99 6.6 x 103 -

I-131 4.6 x 104 2.0 x 101 I-132 6.5 x 103 -

I-133 8.0 x 103 -

I-134 5.3 x 101 -

I-135 1.4 x 103 -

Te-132 1.6 x 103 -

Te-134 1.8 -

Cs-134 5.4 x 104 5.0 x 104 Cs-136 3.7 x 103 3.2 x 101 Cs-137 3.5 x 104 3.5 x 104 Cs-138 4.38 x 101 -

Ba-137m 3.3 x 104 3.2 x 104 Ba-140 1.16 x 102 8.9 x 10-1 La-140 1.2 x 102 1.0 Ce-144 7.4 x 101 6.0 x 101 Pr-144 7.4 x 101 6.0 x 101 Mn-54 1.63 x 102 1.3 x 102 Mn-56 4.1 -

Co-58 2.2 x 103 9.2 x 102 Co-60 2.9 x 102 2.9 x 102 Fe-59 6.1 x 101 1.5 x 101 neglected below 10-1 activity 12.2-60 REVISION 1 - DECEMBER 1989

BRAIDWOOD-UFSAR TABLE 12.2-43 HIGH ACTIVITY SPENT RESIN TANK SHIELDING DESIGN-BASIS RADIONUCLIDE CONTENT (in Curies)

ACTIVITY (Curies) ACTIVITY (Curies)*

ISOTOPE NO DECAY 90-DAY DECAY Br-84 2.0 -

Rb-88 9.8 x 101 -

Rb-89 4.55 -

Sr-89 3.54 x 102 1.03 x 102 Sr-90 5.2 x 101 5.2 x 101 Sr-91 1.7 -

Sr-92 1.8 x 10-1 -

Y-90 3.84 x 101 5.1 x 101 Y-91 5.51 x 101 1.9 x 101 Y-92 4.1 x 10-1 -

Zr-95 9.1 x 101 3.5 x 101 Nb-95 1.3 x 102 6.5 x 101 Mo-99 6.6 x 103 -

I-131 4.6 x 104 2.0 x 101 I-132 6.5 x 103 -

I-133 8.0 x 103 -

I-134 5.3 x 101 -

I-135 1.4 x 103 -

Te-132 1.6 x 103 -

Te-134 1.8 -

Cs-134 5.4 x 104 5.0 x 104 Cs-136 3.7 x 103 3.2 x 101 Cs-137 3.5 x 104 3.5 x 104 Cs-138 4.38 x 101 -

Ba-137m 3.3 x 104 3.2 x 104 Ba-140 1.16 x 102 8.9 x 10-1 La-140 1.2 x 102 1.0 Ce-144 7.4 x 101 6.0 x 101 Pr-144 7.4 x 101 6.0 x 101 Mn-54 1.63 x 102 1.3 x 102 Mn-56 4.1 -

Co-58 2.2 x 103 9.2 x 102 Co-60 2.9 x 102 2.9 x 102 Fe-59 6.1 x 101 1.5 x 101 neglected below 10-1 activity 12.2-60a REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-44 COMPOSITION OF A SINGLE 55-GALLON RADWASTE DRUM FOR SHIELDING ANALYSIS OF DRUM STORAGE AREAS

1. Spent Resin MIXTURE DENSITY VOLUME WEIGHT COMPONENT (lb/ft3) (ft3) (lb)

Radioactive water and spent resins 75 4.5 340 Cement 94 2.85 270 7.35 610 NOTE: Drum composition from the volume reduction system has been intentionally deleted from this table. Braidwood and Byron stations do not intend to use this equipment.

12.2-61 REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE 12.2-45 DESIGN-BASIS SHIELDING SOURCES FOR MAIN AUXILIARY BUILDING CHARCOAL AIR FILTER AND OFF-GAS VENT FILTER*

MAIN AUX. BLDG.

CHARCOAL AIR OFF-GAS VENT ISOTOPE FILTER FILTER**

Br-84 6.4 x 10-8 8.0 x 10-6 I-131 1.8 x 10-3 1.2 x 10-1 I-132 7.1 x 10-6 7.3 x 10-4 I-133 3.0 x 10-4 3.0 x 10-2 I-134 1.7 x 10-6 1.8 x 10-4 I-135 5.3 x 10-5 5.2 x 10-3 Values given are in curies per filter.

Charcoal filters in series are considered to be one filter.

12.2-62

BYRON-UFSAR TABLE 12.2-46 AUXILIARY BUILDING RADIOACTIVE AIRBORNE DESIGN-BASIS CONCENTRATION EXPRESSED IN MPC*

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 330'-0" Auxiliary building Floor Drain Pump Room 2.90-4** 550 2.46-4 1.33-1 1.58-3 Auxiliary Building Floor Drain Sump 3.10-4 400 3.62-5 1.95-2 2.32-4 El. 344'-6 Recycle evaporator room 1.32-3 4300 2.16-3 2.18+0 5.87-2 El. 346'-0 Auxiliary Building Collection Drain Sump Room 2.80-4 1910 6.85-6 4.47-3 4.38-5 Auxiliary Building Equipment Drain Tank Room 2.90-4 1570 8.63-5 4.65-2 5.53-4 Maximum Permissible Concentration, consistent with regulations that were in effect at the time of analysis Read as 2.90x10-4 12.2-63 REVISION 5 - DECEMBER 1994

BRAIDWOOD-UFSAR TABLE 12.2-46 AUXILIARY BUILDING RADIOACTIVE AIRBORNE DESIGN-BASIS CONCENTRATION EXPRESSED IN MPC*

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 330'-0" Auxiliary building Floor Drain Pump Room 2.90-4** 550 2.46-4 1.33-1 1.58-3 Auxiliary Building Floor Drain Sump 3.10-4 400 3.62-5 1.95-2 2.32-4 El. 344'-6 Recycle evaporator room 1.32-3 4300 2.16-3 2.18+0 5.87-2 El. 346'-0 Unit 1 Auxiliary Building Collection Drain Sump Room 2.8-4 1910 6.85-6 4.47-3 4.38-5 Unit 2 Auxiliary Building Collection Drain Sump Room/

Hot Machine Shop 2.8-4 1910 6.85-6 4.47-3 4.38-5 Auxiliary Building Equipment Drain Tank Room 2.90-4 1570 8.63-5 4.65-2 5.53-4

Maximum Permissible Concentration, consistent with regulations that were in effect at the time of analysis

- Read as 2.90x10-4 12.2-64 REVISION 6 - DECEMBER 1996

B/B-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 346'-0" (Cont'd)

Heat Exchanger Valve Aisle 2.20-4 1600 1.70-3 1.73-1 2.05-1 Letdown Chiller Heat Exchanger Room 8.99-5 750 1.49-3 1.52-1 1.79-1 Letdown Reheat Heat Exchanger Room 7.00-5 750 1.16-3 1.18-1 1.40-1 Moderating Heat Exchanger Room 1.10-4 750 1.82-3 1.84-1 2.19-1 Recycle Evaporator Feed Pump Valve Aisle 3.99-5 3200 1.55-4 1.58-2 1.87-4 Recycle Evaporator Feed Pump Room 1.30-4 1600 1.16-3 1.18-1 1.40-3 Recycle Holdup Tank Room -

OA 1.95-4 8000 1.01-4 1.99-2 1.22-4 Recycle Holdup Tank Room -

OB 3.15-4 5750 2.26-4 4.46-2 2.73-4 Regenerative Waste Drain Tank Room 2.70-4 4300 9.77-6 5.20-1 9.52-4 Residual Heat Removal Pump Room 3.50-4 1000 4.33-3 4.94-1 3.13-1 12.2-65 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 346'-0" (Cont'd)

Waste Gas Decay Tank Valve Aisle 1.37-3 17500 2.00-6 5.19-1 1.28-5 Waste Gas Decay Tank Room 9.99-5 4400 2.00-6 6.90-1 1.28-5 El. 355'-4", 358'-2" Waste Gas Decay Tank &

Recycle Evaporator Pipe Tunnel 1.45-3 26100 7.14-4 2.07+0 2.26-2 El. 357'-0" Residual Heat Removal Heat Exchanger Room 1.90-4 1400 1.68-3 1.92-1 1.26-1 El. 364'-0" Auxiliary Building Floor Drain Pump Room 3.40-4 1000 1.59-5 8.58-3 1.02-4 Auxiliary Building Floor Drain Tank Room 9.99-5 750 6.23-6 3.36-3 2.62-4 Blowdown Condenser - Unit 1 4.49-4 4760 1.64-6 - 4.02-6 Blowdown Condenser - Unit 2 4.29-4 2760 2.70-6 - 1.76-5 Centrifugal Charging Pump Room - A 3.99-4 1000 4.95-3 4.60-1 5.99-2 12.2-66 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 364'-0" (Cont'd)

Centrifugal Charging Pump Room - B 3.40-4 750 1.15-2 1.10+0 5.93-1 Chemical Drain Tank Room 9.99-5 1860 1.26-7 6.78-5 8.04-7 Chemical Drain Tank Room 3.40-4 1000 7.94-7 4.29-4 5.08-6 Positive Displacement Charging Pump Room 3.89-4 1000 4.83-3 4.49-1 5.84-2 Chemical/Regeneration Waste Drain Pump Room 2.30-4 1000 2.85-5 2.90-3 3.44-3 Chemical/Regeneration Waste Drain Tank Room 6.99-5 2500 3.47-6 3.53-3 4.20-4 Safety Injection Pump Room - A 3.99-4 1000 5.20-3 5.92-1 3.74-1 Safety Injection Pump Room - B 2.40-4 750 9.89-3 1.06+0 8.11-1 El. 364'-0", 383'-0", 401'-0" Spray Additive Tank &

Pipe Penetration Area 3.99-3 8350 5.93-3 6.05-1 5.25-1 El. 374'-6" Recycle Holdup Tank Pipe Tunnel 1.20-4 2500 2.99-4 5.90-2 3.61-4 12.2-67 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 375'-0" Pipe Tunnel (Q, 15-18) 6.89-4 1250 1.21-2 1.34+0 1.13+0 El. 383'-0" Filter Valve Aisle (M-Q, 11-12) 2.41-3 3360 1.68-3 9.05-1 1.07-2 Filter Valve Aisle (M-P, 13-15) 8.49-4 2010 9.88-4 5.33-1 6.05-3 Filter Pipe Tunnel 1 5.79-4 1760 3.33-3 1.79+0 2.00-2 Filter Pipe Tunnel 2 6.59-4 1600 3.52-3 1.90+0 2.09-2 Filter Pipe Tunnel 3 6.89-4 3810 1.27-3 6.89-1 8.02-3 Heat Exchanger Valve Aisle 2.50-4 1500 2.06-3 2.10-1 2.49-1 Letdown Heat Exchanger Room - A 1.80-4 1300 1.71-3 1.75-1 2.07-1 Letdown Heat Exchanger Room - B 1.60-4 900 2.20-3 2.24-1 2.66-1 Radwaste & Blowdown Mixed Bed Demineralizer Valve Aisle 1.58-3 2500 2.95-4 1.59-1 1.89-3 Blowdown Mixed Bed Demineralizer Cubicle 2.00-4 1500 2.98-4 1.59-1 1.91-3 12.2-67a REVISION 2 - DECEMBER 1990

B/B-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 383'-0" Radwaste Mixed Bed Demineralizer Cubicle 2.00-4 1000 3.89-4 2.11-1 2.49-3 Seal Water Heat Exchanger Room 1.10-4 800 1.70-3 1.73-1 2.05-1 El. 391'-6" Filter Cubicles 5.99-5 260 1.42-3 7.65-1 8.83-3 Filter Cubicles 5.99-5 160 2.55-3 1.37+0 1.63-2 Filter Cubicles 5.99-5 250 1.44-3 7.77-1 8.97-3 El. 394'-6" Auxiliary Steam Pipe Tunnel 1.20-4 2000 - 8.68-2 8.56-1 El. 394'-6" Pipe Tunnel - Unit 1 1.36-3 8000 2.84-3 6.26-1 2.59-1 Pipe Tunnel - Unit 2 1.36-3 10900 2.72-3 8.03-1 2.29-1 El. 401'-0" Boric Acid Tank Room 2.60-4 5800 5.55-4 5.65-2 6.70-2 Main Demineralizer Valve Aisle 1.25-3 8000 3.65-4 1.97-1 2.34-3 Main Demineralizer Cubicles 9.99-5 900 6.24-4 3.37-1 3.99-3 12.2-67b REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 401'-0" (Cont'd)

Main Demineralizer Cubicles 9.99-5 600 7.53-4 4.07-1 4.83-3 Main Demineralizer Cubicles 9.99-5 500 8.31-4 4.50-1 5.33-3 Main Demineralizer Pipe Tunnel 7.99-5 8000 7.09-4 3.83-1 5.75-3 Laundry Drain Tank Room 2.20-4 800 6.43-10 3.47-8 5.14-7 Spent Resin & Concentrates Pump Room 5.99-4 2000 - 1.21-3 8.56-1 Surface Condenser Room - A 1.81-3 5100 - 7.10-2 2.95-2 Surface Condenser Room - B 1.71-3 5700 - 6.00-2 2.48-2 Surface Condenser Room - C 1.56-3 4600 - 6.78-2 2.81-2 El. 414'-0" Radwaste Evaporator Room - A 8.99-4 5100 1.46-4 1.09-1 1.02+0 Radwaste Evaporator Room - B 1.41-3 5700 2.02-4 1.13-1 1.40+0 Radwaste Evaporator Room - C 1.41-3 4600 2.50-4 1.33-1 1.73+0 12.2-67c REVISION 2 - DECEMBER 1990

BYRON-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 417'-0" Concentrates Holding Tank Room 2.00-5 3050 8.12-7 7.93-5 2.31-2 Spent Resin Storage Tank 2.00-5 2450 - - 6.96-2 El. 426'-0 Laundry Room 3.70-4 4250 2.03-10 1.10-7 1.63-7 Volume Control Tank Valve Aisle 7.40-4 2900 1.96-3 1.01+0 1.17-1 Volume Control Tank Room 1.90-4 2900 2.43-3 1.29+0 1.45-1 Waste Gas Analyzer Rack Room(3) 1.40-4 200 1.17-4 5.97+0 2.15-1 Waste Gas Cabinet Aisle (3) 1.80-4 2500 1.26-5 2.15-1 2.29-2 Purge Room 6.20-4 2550 3.01-3 3.07-1 3.63-1 Waste Gas Compressor Room (3) 4.49-4 1000 1.97-4 2.49+0 3.44-1 (1) The leak rates given in the table are based on leakages of 5x10-3 lb/hr per valve or flange; 2x10-2 lb/hr per pump seal for liquid and twice the equivalent liquid volume for gas or vapor. Such large amounts of leakage are expected to be rare, therefore, the actual function of MPC is expected to be a small fraction of the values given.

(2) Following partition factors are used:

Tritium 0.53 for hot liquid, 0.1 for cold liquid Noble 1.0 of all Iodine 0.1 for hot liquid (120F), 0.001 for cold liquid (120F)

(3) Annual average values are reported here. When the gas analyzer is processing gas from recycle evaporator vent condenser, they could exceed the given values temporarily.

12.2-67d REVISION 1 - DECEMBER 1989

BRAIDWOOD-UFSAR TABLE 12.2-46 (Cont'd)

LEAK EXHAUST AIR RATE (1) FLOW RATE NUMBER OF MPC (2)

AREA (gpm) (cfm) TRITIUM NOBLE IODINE El. 417'-0" Low Activity Spent Resin Tank 1.3-4 2750 - - -

Spent Resin Storage Tank 2.00-5 2450 - - 6.96-2 El. 426'-0 Laundry Room 3.70-4 3800 2.27-10 1.23-7 1.82-7 Volume Control Tank Valve Aisle 7.40-4 2900 1.96-3 1.01+0 1.17-1 Volume Control Tank Room 1.90-4 2900 2.43-3 1.29+0 1.45-1 Waste Gas Analyzer Rack Room(3) 1.40-4 200 1.17-4 5.97+0 2.15-1 Waste Gas Cabinet Aisle(3) 1.80-4 2500 1.26-5 2.15-1 2.29-2 Purge Room 6.20-4 2550 3.01-3 3.07-1 3.63-1 Waste Gas Compressor Room(3) 4.49-4 1000 1.97-4 2.49+0 3.44-1 (1) The leak rates given in the table are based on leakages of 5x10-3 lb/hr per valve or flange; 2x10-2 lb/hr per pump seal for liquid and twice the equivalent liquid volume for gas or vapor. Such large amounts of leakage are expected to be rare, therefore, the actual function of MPC is expected to be a small fraction of the values given.

(2) Following partition factors are used:

Tritium 0.53 for hot liquid, 0.1 for cold liquid Noble 1.0 of all Iodine 0.1 for hot liquid (120F), 0.001 for cold liquid (120F)

(3) Annual average values are reported here. When the gas analyzer is processing gas from recycle evaporator vent condenser, they could exceed the given values temporarily.

12.2-67e REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-47 CALCULATED AIRBORNE ACTIVITIES FOR DESIGN-BASIS LEAK RATE IN CONTAINMENT BUILDING TOTAL PRIMARY EXHAUST AIR COOLANT FLOW RATE FRACTION OF MPC**

AREA LEAKAGE* (cfm) IODINES NOBLES H-3 Containment free volume 50 lb/day 3000 1.61-1 1.30+0 1.72-2

Estimate of reactor coolant leakage into the containment atmosphere from valve and pump seals as given in WCAP-8253.

- The partition factor for iodines in a hot liquid (>120F) is 1 x 10-1, H3 partition factor is 0.53 for primary coolant. Use of MPC is consistent with regulations that were in effect at the time of analysis.

12.2-68 REVISION 5 - DECEMBER 1994

B/B-UFSAR TABLE 12.2-48 CALCULATED AIRBORNE ACTIVITIES FOR DESIGN-BASIS LEAK RATE IN RADWASTE BUILDING MAXIMUM EXHAUST AIR LEAKAGE FLOW RATE FRACTION OF MPC*

AREA (gpm) (cfm) IODINES NOBLES H-3 Radwaste Building general 8630 1.80-3 1.50-1 2.80-4 (Estimated) (Estimated) (Estimated)

The partition factor for iodines in a hot liquid (>120F) is 1 x 10-10, in a cold liquid

(<120F) the partition factor is 1 x 10-3; H3 partition factor is 0.53 for hot liquid and 1 x 10-1 for cold liquid. Use of MPC is consistent with regulations that were in effect at the time of analysis.

12.2-69 REVISION 5 - DECEMBER 1994

B/B-UFSAR TABLE 12.2-49 TABLES 12.2-49 THROUGH 12.2-52 AND FIGURE 12.2-1 FOR THE VOLUME REDUCTION SYSTEM HAVE BEEN INTENTIONALLY DELETED.

12.2-70 REVISION 9 - DECEMBER 2002

BRAIDWOOD-UFSAR TABLE 12.2-50 TABLES 12.2-49 THROUGH 12.2-52 AND FIGURE 12.2-1 FOR THE VOLUME REDUCTION SYSTEM HAVE BEEN INTENTIONALLY DELETED.

12.2-71 REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE 12.2-51 TABLES 12.2-49 THROUGH 12.2-52 AND FIGURE 12.2-1 FOR THE VOLUME REDUCTION SYSTEM HAVE BEEN INTENTIONALLY DELETED.

12.2-72 REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE 12.2-52 TABLES 12.2-49 THROUGH 12.2-52 AND FIGURE 12.2-1 FOR THE VOLUME REDUCTION SYSTEM HAVE BEEN INTENTIONALLY DELETED.

12.2-73 REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE 12.2-53 ASSUMED DEMINERALIZER RESIN INVENTORY IN LOW ACTIVITY SPENT RESIN TANK FOR SHIELDING SOURCES CALCULATION VOLUME OF FRACTIONAL COMPONENT RESIN IN CONTRIBUTION TO QUANTITY NAME EACH (ft3) LASRT INVENTORY 4 Blowdown Mixed Bed 113 0.84 Demineralizers 3 Radwaste Mixed Bed 29 0.16 Demineralizers 12.2-74 REVISION 1 - DECEMBER 1989

BRAIDWOOD-UFSAR TABLE 12.2-54 Low Activity Spent Resin Tank Shielding Design-Basis Radionuclide Content (in Curies)

ISOTOPE ACTIVITY (CURIES) ACTIVITY (CURIES)*

Cr-51 8.50-04 8.94-05 Mn-54 8.18-04 6.70-04 Mn-56 3.52-04 --

Co-58 2.57-02 1.06-02 Fe-59 1.04-03 2.57-04 Co-60 8.02-04 7.76-04 Br-84 1.12-03 --

Rb-88 7.72-03 --

Rb-89 2.25-04 --

Sr-89 3.21-02 9.35-03 Sr-90 1.76-03 1.75-03 Y-90 1.31-03 --

Sr-91 8.33-04 --

Sr-92 9.28-05 --

Y-92 9.28-05 --

Zr-95 6.90-04 2.60-04 Nb-95 7.22-04 1.22-04 Mo-99 1.11-02 --

I-131 4.06-01 1.73-04 Xe-131m 3.46-02 1.78-04 Te-132 8.65-01 --

I-132 9.62-01 --

I-133 3.73+00 --

Xe-133m 9.08-02 --

Xe-133 3.09+00 2.10-05 Te-134 9.78-04 --

I-134 2.60-02 --

Cs-134 6.28-04 5.78-04 I-135 6.80-01 --

Xe-135m 1.95-01 --

Xe-135 6.80-01 --

Cs-136 3.22-04 2.81-06 Cs-137 3.22-03 3.20-03 Cs-138 2.09-03 --

Ba-140 3.21-02 2.54-04 La-140 3.05-02 --

Ce-144 3.53-04 --

Pr-144 3.53-04 --

neglected below 10-06 activity 12.2-75 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.2-55 OLD STEAM GENERATOR STORAGE FACILITY SURVEYED DOSE RATES BYRON BRAIDWOOD Inside channel head (middle of tubesheet) 10 R/hr 11 R/hr Inside tube region NA 5 R/hr Outside tube region NA 40 mR/hr Outside steam dome 1 mR/hr 2 mR/hr Notes:

1. The dose rates represent the maximum surveyed dose rates inside and outside the steam generator regions with the steam generator drained.

2. Waste samples at Byron and Braidwood indicate that Co-58 and Co-60 are the dominant gamma-emitting isotopes.

12.2-76 REVISION 7 - DECEMBER 1998

B/B-UFSAR 12.3 RADIATION PROTECTION DESIGN FEATURES Radiation protection design features are provided to reduce direct radiation, control airborne radioactivity, identify radiation areas, decontaminate personnel and equipment, calibrate radiation monitors, and maintain personnel radiation exposure as low as is reasonably achievable (ALARA). Illustrative examples of the application of various radiation protection design features, including several types of shielding to specific components, are provided in Attachment 12.3A.

12.3.1 Description of Facility Design Considerations 12.3.1.1 Equipment Selection, Layout, and Segregation In selecting and shielding equipment and components containing radioactive materials, prime consideration is given to protecting the operating and maintenance personnel from radiation, and to maintain personnel exposures ALARA.

Equipment containing radioactive materials is located in separate rooms or cubicles, where practicable, to protect operating and maintenance personnel from radiation associated with other equipment. Components are remotely operated and/or remotely serviced whenever practicable.

Items which require frequent maintenance and which are radioactive or potentially radioactive, such as pumps, valves, and instrumentation are to the extent practicable, separated from passive radioactive components such as tanks, filters, demineralizers, etc.

Areas containing more than one piece of radioactive equipment are, where practicable, designed and provided with shielding such that maintenance of one item is not restricted by radiation from other pieces of equipment. Where it is not practicable to provide permanent shielding, provisions (discussed in Subsection 12.3.2) for temporary shielding to minimize maintenance doses are provided.

Components which are not radioactive or potentially radioactive are physically separated, to the extent practicable, from components which are radioactive or potentially radioactive.

Radiation detector probe access holes are provided in most shield walls (e.g., shield hatches) for isolated equipment cubicles where access is only by means of removable shield walls.

Partially shielded configurations are reviewed for radiation scattering.

12.3-1

B/B-UFSAR 12.3.1.2 Cubicle Access Access to radioactive or potentially radioactive cubicles or compartments is through entrances designed, where practicable, to permit access to an area of the room which has the lowest or relatively lowest radiation level. Entrances are designed to prevent source radiation from passing directly through entrance openings and into occupied areas. This is done, where practicable, by providing labyrinthine entrances to radioactive and potentially radioactive cubicles.

Typical labyrinthine entrances are shown in Figures 12.3-1 and 12.3-2. Radiation traveling through such labyrinthine entrances collides with the shield walls and consequently can be attenuated to some small fraction of the incident quantity.

Cubicle access for Byron/Braidwood Stations is either through a labyrinthine entrance with an overlap of 1 to 1-1/2 times the passageway width as seen in Figure 12.3-1 or through a double labyrinth arrangement as seen in Figure 12.3-2.

Not all entrances to radioactive areas are designed with labyrinthine entrances. Where labyrinthine entrances are not feasible, other alternatives include:

a. shield doors installed at personnel entrances,

b. removable concrete block walls, and

c. wall and floor removable shield hatches and plugs (such as for the radwaste filter and demineralizer compartments).

The following considerations govern the design of labyrinths:

a. A labyrinth is located and sized to cause unscattered radiation to be attenuated by the required amount of shielding, as shown in Figure 12.3-1. Normally, the labyrinth overlap is designed so that (with worst-case sources) the streaming leaving the labyrinthine entrance due to scattered radiation gives a dose rate which is less than three to five times the design dose rate of the surrounding area.

Where strong sources of low energy gamma radiation are encountered, a double labyrinthine entrance such as depicted in Figure 12.3-2 is used in order to meet this criterion.

b. When the design of the labyrinth is determined by other design considerations, a shield door, isolation of the entrance (e.g., rope off area), extended labyrinth overlap, or a removable labyrinth is also specified.

12.3-2

B/B-UFSAR

c. If the labyrinth height is shorter than the ceiling height, as is often the case, a roof is provided above the labyrinth section.

d. Galleries and other elevated occupied areas are protected from radiation passing through the roof of the labyrinth. The roofs have a thickness which maintain the design dose rate of these elevated areas.

e. Labyrinths inside source cubicles require roofs if any part of a source is higher than the top of the labyrinth. The roof thickness is dependent upon the location of the source, and the thickness is calculated on a cubicle-by-cubicle basis.

12.3.1.3 Draining and Flushing Capability of Equipment Consideration is given in the radiation protection design to identifying the need for adequate draining and flushing capability of equipment designed for radioactive or potentially radioactive service.

The potentially high activity radwaste storage tanks were selected and their designs reviewed to assure adequate draining capability to minimize activity buildup and excessive radiation levels over the plant lifetime. Tanks containing radioactive material have sloped bottoms wherever practicable so that sludge accumulation is minimized and ease of drainage is enhanced.

Where practicable, equipment is selected and the design reviewed to assure that there are no obvious ledges or pockets where radioactivity may be trapped or accumulated.

To the extent practicable, drain piping is of welded construction and is welded in a manner, e.g., using consumable inserts, to minimize crevices which might collect radioactive material. (Use of backing rings in the welds or use of socket welds may be acceptable if the weld is embedded in concrete.)

The design of the spent resin storage and exchange systems is reviewed to assure that the layout and components are such as to prevent the retention of resin beads or fragments in connections, bends, horizontal sections, reducers, etc.

All equipment drains which are considered to be radioactive are directed to appropriate liquid radwaste storage tanks. Sumps are used as intermediate collection points. Such sumps and tanks are appropriately shielded or appropriately located within radiation areas.

The design of the radwaste filters was checked to assure that the filters can be drained and flushed prior to filter element replacement.

12.3-3

B/B-UFSAR Flushing capability of radioactive service equipment is important to assure a minimum of radioactive crud or sludge retention in the equipment prior to maintenance or removal of the equipment.

All potentially high activity source storage vessels were selected and their designs checked to assure adequate draining capability. These tanks include the volume control tank, the spent resin storage tank, the concentrates holding tank, the chemical/regeneration waste drain tank, the auxiliary building floor and equipment drain tanks, and the recycle holdup tanks.

Draining capability is assured:

a. to minimize personnel exposure during testing, surveillance, and maintenance activities and

b. to minimize activity (crud) buildup and avoid excessive radiation levels to accessible areas during plant lifetime.

Adequate draining capability is assured wherever practicable by selecting tanks which have sloped bottoms and which have, or can be provided with, drain lines connected to the lowest level of the tanks. Drainage of the above listed high activity source storage tanks is via remotely operated valves or by valves which are located remotely from the tank cubicle in lower radiation areas. (For location of valves with respect to shielded areas, refer to Subsection 12.3.1.8.)

Flushing of radwaste tanks is accomplished by washing down the tank interiors with demineralized water and/or cleaning agents.

Where practicable, provisions are made to remove crud sedimentation by remote mechanical means with hoses.

Where practicable, flushing of radwaste tank interior is accomplished by an installed sparger (where justified) or by providing a recirculation line for the pump servicing the tank to the bottom of the tank so that a spraying effect can be utilized to get settled deposits in suspension, so that they may be pumped or drained out of the tank. For manual flushing, adequate capability is provided in the form of water connections located near the tank cubicles.

Flushing is required when major maintenance and/or removal of the tank is necessary and also when necessary to reduce radiation levels in adjacent areas due to sources within the tank. Flushed water is directed to tanks having sufficient capacity and shielding necessary to contain and shield the flushed water.

When practicable, the above applies to other high activity source items such as pumps. Where adequate draining and flushing capability is not practicable, shielding is designed to account for worst-case radioactive crud buildup.

12.3-4 REVISION 7 - DECEMBER 1998

B/B-UFSAR 12.3.1.4 Floor and Sink Drains Adequate floor drainage is provided for each room or cubicle housing components which contain, or may contain, radioactive liquids. Floors are properly sloped to the floor drain to facilitate floor drainage and prevent water puddles.

All floor drains which are considered to be radioactive are directed to appropriate liquid radwaste storage tanks. Sumps are used as intermediate collection points. Such sumps and tanks are appropriately shielded or appropriately located within radiation areas. Shielding of radwaste drain piping is discussed in Subsection 12.3.1.6.

To the extent practicable, greater potential radiation area floor drains are segregated from lesser potential radiation area floor drains to protect against backflow of radioactive liquids into lower potential radiation areas, if drainage is blocked or if a large spill occurs. Air circulation through the floor drain system is prevented by the use of water-filled seals (loop seals) or by sealing individual floor drains. The use of such seals also prevents backflow of radioactive gases into the room from the floor drain system.

Sink drains which are expected to contain radioactive fluids are reviewed for appropriate shielding and routing requirements.

Loop seals are present on sink drain lines which may handle radioactive fluids.

All floor drains in the auxiliary, containment, fuel handling, and radwaste/service buildings, except for those areas listed below, are considered to be radioactive and shall discharge to either the auxiliary building floor drain tanks or the chemical drain tank through various sump pumps. Exceptions to this requirement are:

a. diesel-generator oil storage tank rooms,

b. auxiliary feedwater tunnel,

c. main steam/steam generator feedwater tunnel,

d. tendon tunnel,

e. tendon tunnel access area,

f. diesel-generator rooms,

g. cable spreading rooms,

h. switchgear rooms,

i. office areas in service building, 12.3-5

B/B-UFSAR

j. storage rooms in service building,

k. auxiliary electrical equipment room,

l. battery rooms,

m. auxiliary building HVAC equipment area (elevation 451 feet), and

n. essential service water pump rooms.

o. auxiliary building HVAC chilled water coil areas (Byron only) on elevation 451 feet.

p. auxiliary building chiller "A" area on elevation 463 feet.

12.3.1.4.1 Design of Drain System

a. Equipment drains in the turbine building discharge to the two turbine building equipment drain sumps, one per unit, from which they are piped to the turbine building equipment drain tank. At Byron, the drains are treated by the wastewater treatment system and discharged to the circulating water system (CW) flume or to the release tank 0WX26T. At Braidwood, the drains are treated by the wastewater treatment system and discharged to the cooling pond.

b. Equipment drains in the auxiliary, containment, and fuel handling buildings discharge to the two auxiliary building equipment drain collection tanks.

Pumps are provided to pump the drains to the auxiliary building equipment drain tanks.

c. Floor drains that are expected to handle chemical waste solutions from potentially radioactive areas are kept separate from other floor drains and are routed to the chemical drain tank, unless otherwise specified.

d. Leak detection sumps are provided for various areas in the auxiliary building that contain safety-related equipment required for long-term operation.

e. A storm drain system, complete with oil separators, is provided to remove all roof and storm drainage.

f. Borated equipment drains are recycled to the recycle holdup tanks.

12.3-6 REVISION 14 - DECEMBER 2012

B/B-UFSAR

g. High radiation area floor drains are routed separately from low radiation area floor drains to prevent backflow of high contamination into low radiation areas.

h. The top elevation of floor drains are set below nominal elevations of the floor area to be drained.

i. Floors are sloped to the drain to facilitate floor drainage and prevent water puddles.

12.3-6a REVISION 12 - DECEMBER 2008

B/B-UFSAR

j. Slotted cover plates are used to prevent solids from entering floor drain sumps. These cover plates are removable to provide full access to the sump.

k. The arrangement of drains from cubicles containing radioactive equipment is such that air from a zone of high airborne radioactivity potential does not circulate through the drain system to normally accessible areas. The prevention of air circulation is done through the use of loop seals.

l. Drain piping of equipment and systems which carry caustics or acids is the same material as the equipment or system they are draining.

m. Drain lines are sloped 1/8-inch per foot to assure complete drainage of piping. An exception to this is in containment where drain lines may not be sloped 1/8-inch per foot. This does not adversely impact operation of the containment floor drain system, which will continue to function as designed.

n. Drain piping is of welded construction and is welded in a manner to avoid crevices (except where embedded in concrete), which might collect radioactive solids.

All potentially high radioactive drain piping from the equipment to the loop seal is welded using a consumable insert.

o. Equipment drains which interconnect pieces of equipment are designed so as not to inadvertently transfer fluid from one piece of equipment to another.

p. Shielding of radwaste drain piping is provided as necessary. Radwaste drain piping not specifically shielded is routed so that it is not exposed to normally high access areas and general access routes.

Vertical runs of radwaste drain piping not specifically shielded is run against walls and sufficiently isolated so as to facilitate the installation of compensatory shielding, if required.

q. All floor drain piping to the sumps (unless otherwise noted) is carbon steel unless required to be otherwise by design due to flow of corrosive liquids.

r. Primary sample drains are routed to the chemical drain tank and from there processed in the radwaste evaporators (Braidwood only).

12.3-7 REVISION 10 - DECEMBER 2004

B/B-UFSAR 12.3.1.5 Venting of Equipment Where practicable, all radioactive or potentially radioactive equipment (such as filters, demineralizers, and radwaste tanks) is vented to a filtered vent header to minimize the possibility of airborne radioactivity in occupied areas or equipment cubicles due to equipment venting.

12.3-7a REVISION 7 - DECEMBER 1998

B/B-UFSAR Radwaste sumps (i.e., sumps designed to handle drains from radioactive service equipment or from floor areas of potentially radioactive components) are normally either vented to a high radiation area, such as to within the cubicle the sump is located, if it is high radiation cubicle, or to a filtered vent header. Venting of radwaste sumps is important to control the concentrations of radioactive contaminants normally released to the air from potentially contaminated water held in the sumps.

Subsection 12.3.1.5.1 discusses the sumps with venting. For sumps which are in shielded cubicles and which vent to the cubicle, cubicle ventilation rates are such as to assure adequate control over expected airborne concentrations of radioiodine. If venting to other areas is required, the sump covers have air inleakage and have no special provisions for sealing since the sump can maintain a slightly negative pressure with respect to the area in which the sump is located. A small amount of air inleakage to the sump is desirable to maintain air flow through the vent line.

12.3.1.5.1 Sumps Requiring Venting Venting is provided for the auxiliary building equipment drain collection sumps.

Venting of these sumps minimizes the possibility of potential airborne radioactivity in the sump areas. Venting is via a small vent line connected to the sump cover plates. This line is routed to a filtered vent header. Slightly negative pressure is maintained in the vent line with respect to the area in which the sump is located.

12.3.1.6 Routing and Shielding of Lines and Ventilation Ducts 12.3.1.6.1 Routing and Shielding of Lines All potentially radioactive process lines are evaluated to determine proper routing and shielding requirements, based on minimizing radiation exposures to station operating and maintenance personnel.

Radioactive process piping is routed in shielding pipe tunnels, trenches, or chases, or in areas where the radiation field due to the pipe is consistent with the radiation zone for that area.

To aid in preventing crud buildup in process piping, sharp bends, dead ends, and other obvious crud traps are minimized. In general, socket welds and welds employing backing rings are avoided to the extent practicable; these welds contribute to radioactive crud accumulation which results in increased radiation fields near the weld. Where practicable, welds employing consumable inserts are used instead of socket welds or welds using backing rings because the consumable insert weld makes the inside-of-pipe surface smoother and minimizes crevices which may trap crud at the weld. Socket welds and welds employing backing 12.3-8

B/B-UFSAR rings are used, however, if the weld is to be embedded in concrete (such as in concrete floor slabs); for these cases, radiation fields due to radioactive crud accumulation are attenuated by the concrete around the weld.

Shielding of radwaste drain piping (including floor and sink drain piping) is provided as necessary. Radwaste drain piping not specifically shielded is routed so that it is not exposed to normally occupied areas and general access routes. Vertical runs of radwaste drain piping not specifically shielded are run against walls and sufficiently isolated so as to facilitate compensatory shielding, if required.

To the extent practicable, radioactive or potentially radioactive sample lines used for grab samples are routed so that grab samples can be taken in low radiation areas.

Radioactive lines are process system piping, drain lines, sample lines, and other lines which normally do, or may contain, radioactive fluids. Special attention is given to the routing and shielding of radioactive lines.

The following guidelines are followed for routing and shielding of radioactive lines:

a. Routing of radioactive lines in low radiation zones is avoided to the extent practicable.

b. Lines that require shielding are routed in shielded pipe tunnels or in radiation areas to the extent practicable.

c. Penetrations through shielded pipe tunnels are not made by lines which do not, themselves, run through the pipe tunnels.

d. Lines that carry radwaste demineralizer resins, filter backwash, filter/demineralizer sludges, or other particulates have large radius bends and are continuously sloped. On radwaste demineralizer resin lines, welded piping is used but the use of socket welds or welds employing backing rings is avoided to the extent practicable; also, the use of loop seals on these lines is avoided to the extent practicable.

e. Slightly radioactive lines are routed in a manner which minimizes radiation exposure to plant operating and maintenance personnel. Slightly radioactive lines in low radiation zones are, to the extent practicable, routed at a minimum elevation above the finished floor of 10 feet 0 inch, or as high above the floor as is practicable. To the extent practicable, slightly radioactive lines are not routed near 12.3-9

B/B-UFSAR normally traveled passageways, nor near galleries or other elevated work areas.

f. For field routing of 2-inch and under nonseismic radioactive piping, the guidelines listed below are followed.

1. Piping is installed at as high an elevation as is practicable but, in no case, below 10 feet 0 inch from the finished floor level in general access areas, nonsource cubicles, and hallways.

2. Piping is routed as close as possible to existing walls or structures to take advantage of their shielding effect.

3. Radioactive piping is not routed near groups of nonradioactive piping thereby not limiting accessibility to nonradioactive system components.

4. Radioactive piping is not routed near an area radiation monitor thereby causing abnormally high radiation readings which are nonrepresentative of the general area in which the radiation monitor is located.

5. To aid in preventing radioactive crud buildup in the piping, sharp bends, dead ends, and other obvious crud traps are avoided to the extent practicable. The use of socket welds or welds employing backing rings on the piping is avoided to the extent practicable.

12.3.1.6.2 Routing and Shielding of Ventilation Ducts HVAC duct routing was reviewed to assure that air flow is from areas of lower potential radiation contamination to areas of higher potential radiation contamination.

Ventilation duct penetrations of shield walls, floors, and ceilings are evaluated to determine if parapet and labyrinthine shielding around the ducts is necessary. Penetrations in shield walls for HVAC ducts is discussed further in Subsection 12.3.2.3.

12.3.1.7 Waste Filters and Demineralizers The waste filters and demineralizers which accumulate radio-activity and which, if unshielded, could cause the area design dose rate to be exceeded, are located, to the extent practicable, in separately shielded cubicles. Shielding is provided between such adjacent filters and demineralizers to minimize personnel exposure during removal or maintenance operations.

12.3-10

B/B-UFSAR A radiation detector probe access hole is provided in most of the filter and demineralizer removable shield hatches so that radiation levels of the contained equipment may be measured without removing the shield hatches. Figure 12.3-4 shows a typical probe access hole.

The waste filters are designed where practicable to permit removal by a remote handling device. Draining and flushing of radwaste filters is discussed in Subsection 12.3.1.3.

Waste filters also include HVAC filters which may accumulate airborne radioactive materials. These filters are located in areas of the station where access is controlled. Shielding is provided as necessary around HVAC filters (e.g., charcoal filters) to ensure that resultant dose rates from the filter areas are less than the design dose rates for the areas, and to minimize radiation exposure to maintenance personnel during filter removal or maintenance.

HVAC filters are designed for easy removal and sized to allow proper disposal as per Regulatory Guide 1.52, "Design, Testing and Maintenance Criteria for ESF Atmosphere Cleanup System Air Filtration and Adsorption Units of LWRs," Revision 2.

For charcoal air filters, charcoal filtration capacities are such as to assure that radioiodine loadings meet criteria for ESF atmospheric cleanup system air filtration and adsorption units.

12.3.1.8 Valves and Instruments Where practicable, valves are located and shielded from adjacent radiation sources so that they can be operated or serviced without causing excessive exposure to operating or maintenance personnel.

Shielded valve aisles are provided where necessary to allow greater accessibility to frequently operated or maintained valves. The valves are installed in the valve aisle shielded from the equipment they serve. Whether the valves are remotely operated or hand operated, the valves and associated piping are shielded from the valve operating area.

12.3.1.8.1 Valves

a. To extent practicable, all valves servicing radioactive or potentially radioactive equipment are located in shielded valve aisles, apart from the (adjacent) equipment being serviced. Walk-in valve aisles are used where practicable (see Figure 12.3-3). Locating valves in pipe tunnels cannot be avoided entirely, however.

b. All radioactive or potentially radioactive manually operated valves (and associated piping) are shielded 12.3-11 REVISION 1 - DECEMBER 1989

B/B-UFSAR from the valve operating area, to the extent practicable. Where practicable, use is made of remote manual valve operators (valve extensions or reach rods) connected to the manual operated handwheels or geared handwheels and passing through the shielding to allow valve operation in the valve operating area (see Figure 12.3-3). This protects valve operating personnel from radiation due to radioactivity in the valves and associated fluid piping in the valve aisle.

c. Radioactive pipe runs to and from valves located in valve aisles are minimized to reduce the amount of radioactive material in valve aisles. This is done by maximizing the amount of radioactive runs behind shielding (e.g., running as much of the radioactive pipe behind the shield wall which separates the valve aisle from the [adjacent] equipment compartment of the component which the valve services).

d. To the extent practicable, all motor-operated valves and pneumatic operated valves (air-operated valves) which are in radioactive or potentially radioactive service are located in areas which are shielded from the (adjacent) component or item of equipment which the valves service. Locating these valves (which are typically higher maintenance items than manual operated valves) in shielded areas minimizes potential personnel radiation exposures due to other nearby radiation sources during valve maintenance and inservice inspection.

e. Valves servicing radioactive or potentially radioactive equipment are installed and positioned with respect to other valves so that (1) service or maintenance time is minimized, and (2) compensatory shielding (e.g., lead blankets) is used, where practicable, to protect workers from adjacent radioactive valves and piping.

f. For valve maintenance, provision is made for draining or flushing the valve and associated connecting lines of radioactive fluids so that radiation exposures are minimized.

Figure 12.3-3 shows a typical walk-in valve aisle arrangement for Byron/Braidwood Stations.

12.3.1.8.2 Instruments

a. Output devices such as instrument readouts, pressure switches, electrical bistable devices, electric converters, control devices, etc., are located and positioned in areas (e.g., at valve operating stations) which result in the lowest personnel 12.3-12

B/B-UFSAR exposures, consistent with other requirements such as instrument accuracy and precision. Use of transducers is maximized in high radiation areas.

b. The following is considered in the location and positioning of the instrument readout devices to assure ALARA exposures.

1. Locate in readily accessible areas.

2. Position at convenient elevation for observation and application of parallax corrective devices.

3. Face readout toward direction convenient for reading.

4. Provide easily readable numbers and easily observable pointers and needles.

5. Preclude or minimize application of scale multipliers on readout.

6. Locate to take advantage of amount of lighting available.

7. Locate instruments and instrument readouts away from local hot spots caused by streaming radiation or from the accumulation of radioactivity in lines, ducts, filters, and equipment.

c. Wherever practicable, radiation monitoring equipment with remote readout is located in areas to which personnel normally have access.

12.3.1.9 Contamination Control and Decontamination In addition to the safety design features discussed above, the following safety design features specifically relating to decontamination and contamination control are incorporated into the radiation protection design of the station.

a. Curbs Where practicable and where failure of radioactive storage tanks, vessels, or associated piping is postulated, either the floor of the cubicle is situated at an elevation lower than the entrance to the cubicle or curb walls are provided to restrict radioactive material to the cubicle.

Curbs are provided for equipment decontamination pads to restrict washdown water to the pad and avoid contamination of adjacent areas.

12.3-13

B/B-UFSAR

b. Protective Surface Coatings Wherever there exists a potential for leakage or spillage of radioactive material onto concrete surfaces (e.g., shield walls, floors, or ceilings),

such surfaces are coated with a nonporous coating to enhance decontamination.

The following guidelines and criteria are used for the application of coating systems to potentially contaminated concrete surfaces in the station to enable them to be effectively decontaminated.

The function of the protective coating system is to facilitate decontamination of surfaces by providing a clean, smooth, and hard finish that is minimally free of cracks, is nonabsorbent, and is water-repellent. Surface contaminants can then be removed by means of washing, sweeping, scrubbing, or wiping in one or more applications.

a. The coating systems are capable of performing their surface protective functions throughout the 40-year plant lifetime (including reasonable maintenance and touch-up activities) and under the variable radiation source and environmental conditions anticipated for the plant.

b. The coating systems applied to floors, curbs, dado, and wainscot are capable of maintaining their integrity in protecting these surfaces under conditions of water immersion. The coating systems used on floors, curbs, and dado is therefore solvent-based. The wainscot can be either solvent or water-based.

c. To enable the coating systems to perform their intended function, a surface preparation system appropriate to the surface as well as to each coating system, is first applied. The surface preparation system includes surface cleaning, the filling of holes and the application of primer coating.

d. The coating systems used on floors and ramps are capable of maintaining their integrity in protecting these surfaces under the traffic patterns (people, lift trucks, etc.) anticipated in the various areas.

The thickest of the field coating systems should be specified for such areas that involve continuous use to avoid deteriorating and thereby compromising the coating.

Protective coating systems are applied to concrete surfaces on the following basis:

12.3-14

B/B-UFSAR

a. Where no other requirements are necessitated, all walls are coated to 1 foot-0 inch dado height to protect this lowest section during sweeping and washing of the floor.

b. Walls that require only partial height coverage are coated to one of several standard wainscot heights (usually 5 feet-0 inch or 8 feet-0 inch). General examples include the walls of potentially radioactive heat exchangers, and certain access area locations.

c. Cubicles containing radioactive equipment that reach above the highest wainscot level are coated to full height and in most cases, the ceiling. The coating of rooms utilizing monorail or crane systems for handling radioactive materials are based on the elevated height of the materials.

d. Cubicles containing radioactive processing equipment such as pumps or pressurized pipe with valving are coated to full height, where necessary. Potentially radioactive contaminated water can come from the room (or area) on the floor above, through penetrations in the ceiling.

e. Walls are coated to full height if the potential exists for leakage of radioactive contaminated water from the room or area on the floor above, through penetrations in the ceiling.

f. Cubicles, rooms, and areas that require complete wall coverage have their ceiling fully coated as well.

This includes the underside of removable shield hatches and plugs as well as fixed ceilings.

g. Pipe tunnels that are accessible and contain radioactive pipes are fully coated.

Areas that require partial wall coverage (although complete wall coverage may be dictated by equipment size) include the following:

a. areas around sampling stations or panels receiving radioactive process streams for monitoring;

b. areas through which heavy traffic patterns are expected; and

c. clothing change areas, personnel monitoring points, and counting room.

12.3.1.9.1 Equipment Decontamination Facilities Equipment decontamination facilities are provided in the station as required for the decontamination of contaminated equipment, 12.3-15

B/B-UFSAR tools, etc. The design of these facilities includes adequate shielding, and ventilation and filtration of the room air.

A separate area, the equipment decontamination facility, is provided on elevation 346 feet 0 inch in the auxiliary building for cleaning, and decontaminating tools and small pieces of equipment.

12.3.1.9.2 Personnel Decontamination Facilities A personnel decontamination facility is supplied on elevation 426 feet 0 inch in the auxiliary building to provide for prompt decontamination of plant personnel, if the need should arise.

12.3.1.9.3 Station Decontamination Radiation decontamination of the station is currently expected to be required at least once during the life of the station. The radiation protection safety design features discussed above assure less complicated station radiation decontamination when it is required.

12.3.1.10 Traffic Patterns and Access Control Points Traffic patterns are established to maintain occupational radiation exposures ALARA. Anticipated traffic patterns have been used to determine design dose rates in the various areas, and thus have affected the determination of radiation zones.

The majority of normal personnel traffic occurs between the service building and the auxiliary building. The remainder of the traffic occurs in operating areas (where panels and motor-control centers are located), hallways, elevators, and stairwells.

Access control points (i.e., check points for personnel) are flexible and are determined on a day-to-day basis, depending on contamination levels and maintenance activities.

12.3.1.11 Radiation Zones Radiation zones have been defined as a means of classifying the occupancy restrictions on various areas within the plant boundary. The design criteria for each zone are described in the subsections which follow and are tabulated in Table 12.3-1. The radiation zones assigned to the areas of the plant, and upon which the shielding has been designed, are shown in the radiation zone maps in Figures 12.3-27 through 12.3-71.

Selection of appropriate design dose rates for particular plant areas is one method of maintaining individual doses within regulatory limits. Maintaining total collective exposure ALARA 12.3-16 REVISION 8 - DECEMBER 2000

B/B-UFSAR has been considered in plant layout and zoning designations. The plant design has an abundance of general access areas. These areas are designed so that 100% occupancy in these areas results in a total annual dose which is far below the regulatory limits.

The general access areas are an integral part of the ALARA concept of exposure control. These areas are used to travel from one part of the station to almost every other part of the station. If it becomes necessary, a limited amount of maintenance can be performed in some sections of the general access areas, but such sections are used only when reductions in exposures result.

The zone designations are only a tool to aid administrative controls. The zones given in Table 12.3-1 are based on design-basis radiation sources. The actual maximum dose rates for the zones that are less than or equal to 100 mrem/hr are expected to be a small fraction of the given dose rates. A more precise zoning is obtained by using the data from periodic radiation surveys. The dose rates were determined using limits that were in effect at the time when the zones were designated.

Zoning decisions for radiation areas during operation are determined by the residual radioactivity of the equipment (due to plateout and crud buildup) and the activity of material which may be present in the equipment. The shutdown condition for the same areas has a much lower level of radiation because only the residual activity is present. The majority of the occupational radiation exposure (~80%) is accumulated while performing surveys, inspections, and maintenance during operation and/or shutdown conditions.

The total person-rem exposure during surveys and inspection is kept ALARA administratively. The administrative controls are described in Subsection 12.5.3.1. The station design aids the administrative controls by segregating equipment with shielding, which allows high maintenance items to be located in ALARA radiation environments. Thus, the background radiation due to one type of equipment on a second type is kept to a small fraction of the residual radiation of the second type. Since two or more of the same type of equipment can be in the same area, administrative controls will determine the maintenance procedure necessary to keep radiation exposure ALARA.

Zone I-A Zone I-A has no restriction on occupancy. A I-A area would represent, for example, plant site where radiation due to occupancy on a 40 hr/week, 50 week/yr basis, will not exceed the whole body dose of 0.5 rem/yr. The environs around the plant such as the pump house, electrical switchyards, and turbine hall, are examples of a Zone I-A area.

It is expected that nonplant personnel or visitors to the site will receive considerably less than 0.5 rem/yr because of the relatively small time interval during which they are on the site.

12.3-17 REVISION 8 - DECEMBER 2000

B/B-UFSAR Zones I-B, I-C, and I-D Zones I-B, I-C, and I-D are areas which individuals can occupy on a 40 hr/week, 50 week/yr basis, and not exceed a whole body dose of 1.25 rems per calendar quarter. The design dose rates are from 0.5 to 2 mrem/hr in these zones. The area will remain accessible. Corridors in the auxiliary building and areas outside radioactive enclosures where personnel can walk freely are included in this zone.

Zone II-A Zone II-A is a radiation area that plant personnel can occupy periodically. This zone has a design dose rate of 4 mrem/hr.

The radiation level in a Zone II-A area will be posted, but the area will remain accessible to the plant personnel.

Zone II-B, II-C, and II-D Zones II-B, II-C, and II-D are areas where dose rates are from greater than 4 mR/hr to 100 mR/hr. Occupancy is limited. The time a worker with a permit can stay in this room is determined by four factors:

a. the actual radiation level in the room;

b. the nature of the radioactivity (airborne, gamma, etc.);

c. the past radiation history of the worker; and

d. nature of the required job.

The "nature of the required job" means that the necessity of the job being done to ensure the safe operation of the station will be considered when work in these radiation areas is being planned.

Auxiliary equipment which requires manual operation or inspection or maintenance during unit operation will not be located in these zones.

All equipment in areas designated as Zone I-A, I-B, I-C, I-D, or II-A will not contain radioactive materials, or if it does, the activity will be such that the dose rate outside the equipment is consistent with the design dose rate in the area. Such equipment could include fluid system, monitor tanks, and monitor pumps.

Zone III Zone III represents areas where design dose rates are in excess of 100 mR/hr and occupancy periods are limited.

12.3-18 REVISION 8 - DECEMBER 2000

B/B-UFSAR Zone IV This zone is not assigned at Byron/Braidwood.

Zone V Zone V is the main control room area. Zone V normal dose rate will be less than or equal to 0.2 mR/hr during normal operations.

During an accident, the integrated whole body dose will not exceed 5 rem.

12.3.1.12 Laboratory Complex The station laboratory complex is located in a controlled access area on the mezzanine floor of the auxiliary building. The facilities located in this complex are: a high level laboratory, a low level laboratory, a counting room, mask cleaning room (Byron only), instrument storage room (Braidwood only), personnel decontamination room, chemistry offices, and supervisor offices.

This complex serves as a center for the chemistry activities at the station.

12.3.1.12.1 High Level Laboratory The high level laboratory is designed to provide for the safe and efficient processing and analysis of radioactive and potentially radioactive samples. Such samples may be expected for such systems as the: primary coolant loop, chemical and volume control, fuel handling and storage, steam generator blowdown, and radwaste.

The major facilities provided in this laboratory are: fume hoods (with HEPA filtered exhausts), sinks (with drains routed to the liquid radwaste system), sufficient workbench space to allow frequently used equipment to be left in place, sufficient built-in storage space to assure a safe, uncluttered work environment, computer grade regulated electrical circuits, and a close tolerance HVAC system (temperature and humidity) to assure optimal performance of sensitive laboratory equipment.

To minimize the accumulation and spread of surface contamination; floor coatings, surface coatings, workbench surfaces, fume hood interiors, and sink and drain pipe materials are chosen to minimize the adherence and ease the removal of contamination. To minimize the spread of airborne radioactivity, fume hoods are provided for the storage and processing of volatile radioactive samples, and the high level laboratory is kept at a negative pressure with respect to the adjacent low level laboratory, counting room and laboratory corridor. All air exhaust from this laboratory is filtered prior to its release to the environment via the station vent stack.

12.3-19 REVISION 8 - DECEMBER 2000

B/B-UFSAR 12.3.1.12.2 Low Level Laboratory The low level laboratory is located adjacent to the high level laboratory on the mezzanine floor of the auxiliary building. It is designed to provide a radiation and contamination free environment for the chemical preparation and analysis of nonradioactive samples (i.e., those samples which could not pose a radiological danger to the laboratory workers). The major equipment provided in the low level laboratory includes: fume hoods, sinks, workbenches, and storage facilities.

12.3.1.12.3 Counting Room The counting room is located near the high and the low level laboratories on the mezzanine floor of the auxiliary building.

This room is provided with computer grade regulated electric circuits and nonfluorescent lighting to assure the optimal performance of the counting equipment. The desired radiation level in the counting room should be below background. To assure that the counting room will not be affected by any in-plant airborne radioactivity the room is maintained at a positive pressure with respect to all surrounding areas and is ventilated with fresh filtered and conditioned air. The room HVAC is designed to maintain the temperature and humidity tolerances required by the detectors and their associated electronics and computer equipment. The use of thick concrete walls for shielding the counting room was precluded due to considerations of natural radiation emanating from the concrete itself.

The original equipment provided in the counting room includes:

a. gamma-ray spectrometer subsystem,

b. multichannel analyzer subsystem,

c. data analysis subsystem,

d. standard alpha, beta counting subsystem,

e. low-background alpha, beta counting subsystem,

f. automatic sample changer (for d and e), and

g. liquid scintillation counting system.

Localized radiation shielding is provided for the counting equipment when needed.

12.3.1.12.4 Chemistry Storage A chemistry storage room is located in the general area of the low level chemistry room and the counting room on the mezzanine 12.3-20 REVISION 8 - DECEMBER 2000

B/B-UFSAR THIS PAGE WAS INTENTIALLY DELETED.

12.3-21 REVISION 16 - DECEMBER 2016

BYRON-UFSAR floor of the auxiliary building. It provides storage space for the chemistry supplies used within the laboratory complex.

12.3.1.12.5 Mask Cleaning Room The laboratory complex includes a mask cleaning room. The room includes space and equipment for collecting, cleaning, inspecting, and storing respiratory protective equipment.

12.3.1.12.6 Personnel Decontamination Room The decontamination room associated with the complex is designed to facilitate the decontamination of station personnel. A shower and sink are provided.

12.3.1.12.7 Office Space The chemistry office and the supervisor offices in the laboratory complex are provided to assure adequate, local office space for the laboratory complex workers.

12.3.1.13 Laundry Facility The station laundry facility is located on the mezzanine floor of the auxiliary building. It is designed to receive, store, and distribute the radiological protective clothing used in-plant.

The floor and surface coatings in the laundry have been chosen to minimize the buildup and ease the removal of surface contamination. The laundry room is kept at negative pressure with respect to all surrounding areas to minimize the spread of airborne contamination originating from the handling of contaminated equipment.

12.3-21a REVISION 10 - DECEMBER 2004

BRAIDWOOD-UFSAR floor of the auxiliary building. It provides storage space for the chemistry supplies used within the laboratory complex.

12.3.1.12.5 Instrument Storage Room An instrument storage room is located within the laboratory complex.

12.3.1.12.6 Personnel Decontamination Room The decontamination room associated with the complex is designed to facilitate the decontamination of station personnel. A shower and sink are provided.

12.3.1.12.7 Office Space The chemistry office and the supervisor offices in the laboratory complex are provided to assure adequate, local office space for the laboratory complex workers.

12.3.1.13 Laundry Facility The station laundry room is located on the mezzanine floor of the auxiliary building. It is designed for storage of Radiation Protection equipment and supplies, sorting of low level radioactive trash, and occasional laundering of contaminated personal clothing. The floor and surface coatings in the laundry room have been chosen to minimize the buildup and ease the removal of surface contamination. The laundry room is kept at negative pressure with respect to all surrounding areas to minimize the spread of airborne contamination originating from the handling of contaminated equipment.

12.3-21b REVISION 16 - DECEMBER 2016

B/B-UFSAR The radiation zone map in Figure 12.3-31 shows the laundry room to be in a zone of less than or equal to 4 mrem/hr. This is under the most extreme conditions. The laundry room dose rates should average less than 1 mrem/hr during normal operation.

12.3-21c REVISION 10 - DECEMBER 2004

B/B-UFSAR 12.3.1.14 Survey Instrument Calibration Room The instrument calibration room located on the Unit 1 side of the auxiliary building on the 401-foot level is designed to provide a location where radiation protection instrumentation can be calibrated, stored, serviced, and decontaminated when necessary.

12.3.1.15 Locker Room Facilities Change areas are provided in the plant as necessary for individuals to don protective clothing for work in contaminated areas. Storage of personnel clothing is provided at these locations or other designated areas.

12.3.1.16 Design Features to Assist Decommissioning The radiation protection design features established for station operation will aid in maintaining occupational radiation exposure ALARA during decommissioning. The shielding design allows for efficient mothballing and entombment. Decommissioning by removal of all contaminated and activated equipment will be aided by remote handling of equipment, equipment layout (Subsection 12.3.1.1), and administrative planning, which includes the health physics program (Section 12.5).

Specifications and limitations on cobalt and nickel content in equipment components will serve to limit radiation doses from the buildup, transport, and deposition of activated corrosion products in reactor coolant and auxiliary systems during both operation and subsequent decommissioning. A summary of the features in Westinghouse PWRs that reduce occupational exposure are given in Reference 9. Information on the steps taken to minimize Co-58 and Co-60 is given in Chapter 6 of Reference 9.

Radiocobalt and crud buildup in the primary coolant above 250F are controlled below specification limits by continuous monitoring and controlling of the oxygen concentration.

Hydrazine additions to the primary coolant and a hydrogen or nitrogen blanket in the volume control tank are the means of oxygen control. Control of pH in the primary coolant is accomplished by lithium hydroxide addition and is maintained between a pH of 4.5 12.3-22 REVISION 10 - DECEMBER 2004

B/B-UFSAR and 10.5 depending on the disassociation of the boric acid present in the primary coolant.

The National Environmental Studies Project of the Atomic Industrial Forum has analyzed the decommissioning alternatives for LWRs in Reference 10. The majority of the estimated PWR occupational radiation exposure due to removal/dismantling comes from decontaminating the primary and radwaste systems.

Experience gained through decontamination of Exelon Generation Company stations will be applied to decommissioning, which should produce additional dose saving procedures. Estimated occupational radiation exposures from the study are given in Table 12.3-7. The dominant radioactive isotopes that are expected to be found during decommissioning are given in Table 12.3-8.

12.3.1.17 Old Steam Generator Storage Facility Features Four old Unit 1 steam generators are stored in the old steam generator storage facility (OSGSF). The OSGSF has an 18-inch concrete roof and 30-inch concrete walls. A vestibule, which contains a lockable, personnel-access door, is designed to minimize radiation streaming beyond the outside surface of the OSGSF. The OSGSF has a water collection sump. The sump access and monitoring port are located within the vestibule and are designed to allow monitoring of the collection sump without entry into the facility (entry only into the vestibule is required) and to allow radiological survey access. The sump is checked for water content in accordance with station radiation protection procedures, as well as sampled and discharged in accordance with applicable station procedures. The general arrangement of the OSGSF is given in Drawing M-24-23.

The OSGSF has been designed such that the dose rates at the exterior of the facility (walls and roof) are within the dose limits of 10 CFR 20. The area exterior to the OSGSF is a Zone 1-A area. The radiation zones assigned to the OSGSF are shown in Figure 12.3-71.

12.3.2 Shielding The design of the station shielding is based on the design dose rates and the established design criteria. Using the sources given in Section 12.2 and the shielding design criteria, the shielding design is determined.

The original licensed power level was 3411 MWt. The original source term and shielding analyses were performed at a power level of 3565 MWt. Byron and Braidwood Nuclear Stations have uprated the core power level twice. First to a core power level of 3586.6 MWt, then to the Measurement Uncertainty Recapture uprate core power level of 3645 MWt. Accounting for core power level uncertainty, the analyzed core power level is 3658.3 MWt.

This represents an increase of 2.6% from the original design basis. The core fission source given in Table 12.3-5 will increase by 2.6% after the uprates.

12.3-23 REVISION 15 - DECEMBER 2014

B/B-UFSAR As stated in Section 12.2.4, the plant design basis radiation source terms will either remain valid for uprate or will increase by a maximum of 0.6%. As noted in that section, this small percentage increase is well within the conservative margin that was maintained in calculating the original source terms and modeling the shielding configurations to develop the design dose rates. Consequently, the radiation protection design features described in this section remain valid for power uprate.

Note that during plant operations, the plant ALARA program confirms adequacy of shielding and maintains the radiation levels in the plant within the design limits of the normal operation plant radiation zones.

12.3.2.1 General Shielding Design Criteria Every component that handles radioactive fluids may require shielding; the thickness of which is based on the operational cycle of the component, the design dose rate, and the shielding material.

12.3.2.1.1 Regulatory Requirements The shielding design dose rates for Byron/Braidwood meet 10 CFR 20 and 10 CFR 50, which are concerned with allowable radiation to individuals in restricted and unrestricted areas. The only shielding required to be safety-related is the control room and the primary containment shielding; this shielding satisfies the requirements stated in Criterion 19 of 10 CFR 50, Appendix A, and 10 CFR 20.

12.3.2.1.2 Shielding Requirements Radiation protection of personnel, equipment, and materials is largely dependent upon the adequacy of the design of the station shielding system. Radiation shielding has the passive protection function of radiation attenuation and consists of material placed between radiation sources and personnel and/or equipment and materials needing protection from radiation.

The shielding system is designed and constructed to assure that the station can be operated and maintained such that the resultant radiation level and doses are within the limitations of applicable regulations and are as low as is reasonably achievable (ALARA). Specific design dose rate limits recommended to achieve 12.3-23a REVISION 9 - DECEMBER 2002

B/B-UFSAR this objective are discussed in Subsection 12.3.1 and listed in Table 12.3-2.

Shielding must be capable of performing its protective function throughout the plant lifetime and under the variable source and environmental conditions associated with all normal, anticipated abnormal operational, and design-basis accident conditions identified in the safety analysis reports and as noted in this section.

a. Normal Operating Conditions For the purposes of shielding design, normal station operating conditions are considered to include conditions generally known as anticipated abnormal operational occurrences. Two modes of normal station operation are:

1. normal power operation of the reactor, including anticipated operational occurrences, and

2. normal shutdown of the reactor.

Shielding is designed to provide the required protective function under such conditions.

b. Accident Conditions Station shielding provides protection to plant operating personnel and the general public under postulated design-basis accident conditions as defined in the Chapter 15.0.

1. Control Room Habitability The main control room and associated areas are shielded such that, after a postulated design basis accident, the dose in the control room for the duration of the accident will not exceed 5 rem TEDE, including ingress and egress, as per requirements 10 CFR 50, Appendix A, Criterion 19.

Subsection 6.4.2.5 describes control room shielding.

The radiation shielding protecting the main control room (and associated areas) is designed based on the anticipated radiation environment resulting from a postulated design basis accident. Figure 6.4-2 shows an isometric view of the main control room shielding.

12.3-24 REVISION 12 DECEMBER 2008

B/B-UFSAR

2. Direct Offsite Doses All sources in the plant are adequately shielded to assure that radiation levels at the restricted area boundary are in compliance with 10 CFR 20 limits. Adequate station shielding is provided to limit site boundary doses, due to direct and scattered radiation from contained sources within the plant, practicing ALARA during normal operation in conformation with 10 CFR 20 and to within the limits specified in 10 CFR 100 for accidents analyzed using TID-14844 or 10 CFR 50.67 for accidents analyzed using alternative source term methodology.

3. Seismic and Safety Classification Structural walls of the station are designed, as required, to meet Seismic Category I requirements. Walls which are shielding walls may be designed Seismic Category I, depending upon the particular design requirements other than radiation protection requirements (e.g.,

structural integrity, load bearing capacity, etc.) that the walls must meet.

The primary shield, the shield walls for the main control room, and the shield walls for the spent fuel pool are examples of shield walls which are designed Seismic Category I.

c. Protection of Equipment Appropriate shielding is provided, where needed:

1. to limit radiation heating of building structural concrete,

2. to reduce neutron activation of equipment, and

3. to limit radiation to equipment and materials.

Protection from neutrons and from neutron-induced gamma rays is important around neutron sources such as the nuclear reactor core. The primary shield around the reactor vessel is an example of station shielding designed to protect personnel and equipment against neutron radiation and neutron-induced gamma rays.

d. Additional Requirements In addition to the radiation protection functions discussed above, the shielding systems have other 12.3-25 REVISION 12 - DECEMBER 2008

B/B-UFSAR functional requirements. These generally depend on the location of the shield and the access requirements to equipment or areas beyond the shield. Thus, access to an area may be through the shield itself; e.g.,

through removable shield walls. Removable shield walls, portable shields, and compensatory shielding are discussed in Subsection 12.3.2.1.6.

12.3.2.1.3 Design Requirements The station shielding system must be capable of performing its protective functions throughout the plant lifetime and under the variable source and environmental conditions which are anticipated and/or postulated for the plant.

The radiation attenuating materials which comprise the station shielding system are selected to assure no significant loss in radiation attenuation characteristics for at least 40 years of plant operation.

12.3.2.1.4 General Description and Design Parameters The shielding system includes all concrete walls and associated radiation attenuating materials (e.g., lead, steel, and water) which are used to protect the public, plant personnel, equipment, and materials from radiation emitted from radioactive sources contained or generated within the plant. The radiation exposure of individuals, equipment, and materials is a function of the following basic parameters, which are given due consideration in the shielding design:

a. source strength (type, intensity, energy);

b. number of sources, source geometry, and self absorption factors;

c. shielding material, geometry, and mass between source(s) and receptor;

d. distance between source(s) and receptor;

e. time that receptor is exposed; and

f. allowed dose rate or dose.

Where radioactive crud buildup sources are known, the source strength parameter is appropriately adjusted and the shielding designed to accommodate the effects of crud buildup for at least 10 years of reactor operation. Where radioactive crud buildup sources are not known, but expected, the shielding design reflects appropriate conservatism to accommodate the expected effects of crud buildup for at least 10 years of reactor operation, and/or protective measures are used, where practicable, e.g., those discussed in Subsection 12.3.1.

12.3-26

B/B-UFSAR 12.3.2.1.5 Shielding Materials and Construction Methods Bulk shielding structures such as cubicle shielding walls, floors, and ceilings are mainly designed of ordinary concrete, either of (solid) block or poured-in-place construction. Where space limitations are encountered, a special high density concrete (e.g., Hematite concrete) is employed to assure adequate radiation protection. Concrete is a mixture of materials, the exact proportions of which may differ from application to application. Concrete for radiation shielding is classified as ordinary or high density according to the unit weight of the aggregate. The design of concrete mixtures and forms, the construction of concrete radiation shielding structures, and the quality assurance provisions needed to verify that the desired quality of construction has been achieved is in accordance with accepted design criteria for concrete radiation shields.

Poured-in-place concrete construction is normally used for shielding structures which are load-bearing structural walls.

Concrete block walls are provided where necessary to accommodate equipment installation, removal, and construction. Concrete block wall installation is controlled to assure as-built radiation attenuation characteristics similar to those expected from equivalent poured concrete.

In the case of the primary shield around the reactor vessel, nuclear heating is not severe enough to warrant special designs (e.g., water cooling coils) for cooling the primary shield.

The reactor vessel nozzle inspection cavity hatches are made of stainless steel with no special neutron shielding material. They do not exhibit neutron shielding qualities.

Where a potential of leakage or spillage of radioactive material exists, effective features are provided in the design of the shielding to prevent the spread of contamination by seepage through walls. As discussed in Subsection 12.3.1.9, wall surfaces are coated with a nonporous coating to permit effective decontamination.

12.3.2.1.6 Removable Shield Walls, Portable Shielding, and Compensatory Shielding Shielding is designed to be removable, where required, to provide personnel access for inspection, servicing, maintenance, or replacement of plant equipment.

Removable shield panels are provided in shield walls, floors, or ceilings as necessary where frequent access for maintenance or removal of equipment is required and if radiation levels in the 12.3-27 REVISION 9 - DECEMBER 2002

B/B-UFSAR area can cause excessive exposure. Such shielding is designed to minimize exposure to operating and maintenance personnel.

Compensatory, portable, or temporary shielding is considered in station design only as required where other more permanent shielding is not practicable. Where compensatory shielding is necessary, provisions are made to accommodate such shielding in terms of space, structural loading, clearances, and equipment accessibility.

The station shielding system uses three types of removable shield walls: stacked unmortared block, shield hatches and plugs, and shield doors. The primary functions of a removable wall are equipment installation, inspection, maintenance, and removal.

The following are guidelines for the design of removable shield walls. Note: the term major maintenance requires the removal of a removable shield wall in addition to repairing and maintaining equipment.

12.3.2.1.6.1 Stacked (Unmortared) Block Removable stacked block walls that are provided to accommodate removal of equipment are constructed such that the top of the removable unmortared block sections are offset and provided with a lintel arrangement. The blocks are held in place by special metal frames to resist lateral pressure and seismic loads. Use of stacked unmortared block avoids unnecessary exposure associated with disassembly or mortared blocks.

Removable stacked block shield walls are used in the shield design when a room contains equipment that seldom requires replacement or major maintenance. Seldom is defined in the section as once a year. The type of shielded equipment which fits into this category are heat exchangers, pumps, and radwaste tanks.

12.3.2.1.6.2 Removable Shield Hatches and Plugs Removable shield hatches (or removable floor slabs) and plugs are used in the shield design when a room contains equipment which often requires replacement or maintenance. Often is defined in this section to mean more frequent than once a year.

In addition to equipment that requires frequent maintenance, shield hatches or plugs are used, whenever practicable, for access to equipment and piping which have, or are in radiation areas that have, a dose rate greater than 3 R/hr.

The use of removable shield hatches or plugs minimizes the maintenance exposure to station personnel; shield hatch and plug design and construction shall be in accordance with ANSI N 101.6-1972.

12.3-28 REVISION 7 - DECEMBER 1998

B/B-UFSAR A radiation detector probe access hole is provided in most of the filter and demineralizer removable shield hatches so that radiation levels of the contained equipment may be measured without removing the shield hatches. This is provided by boring a vertical stepped hole in the top of the shield hatch for insertion of a radiation detector. The arrangement is pictured in Figure 12.3-4.

The types of equipment that require removable shield hatches are demineralizers, filters, and pumps and motors, which are radioactive or are in radioactive areas.

12.3.2.1.6.3 Shield Doors Shield doors are used when access requirements, maintenance requirements, or design consideration make it undesirable to adequately employ the removable shield walls mentioned previously. Shield doors can also be used with labyrinthine entrances where the dose rate at the entrance due to scattered radiation is greater than the design dose rate.

12.3.2.1.7 Inspection (Inservice) and Maintenance Requirements Shielding is designed to permit access for required inspections, testing, and maintenance of plant systems and components which require these functions.

During construction, shield walls are visually inspected for cracks and separations that might compromise the shield. There are initial preoperational radiation surveys taken as well as periodic routine radiation surveys during power operation. These surveys serve as a check on the radiation buildup within auxiliary equipment and the adequacy of shielding design.

Installed radiation monitoring systems survey continuously the radiation condition at certain areas of the plant and also serve as a check on the adequacy of shield wall design and construction.

As discussed in Subsection 12.3.1, biological protection of personnel during anticipated inspection and maintenance activities are considered in shielding design in the effort to maintain exposures ALARA.

12.3.2.1.8 Shield Thicknesses Shield thicknesses are designed to reduce the average area dose rate to or below the assigned area dose rate level for worst-case conditions of normal plant operation or, where applicable, for accident conditions. Worst-case conditions include source terms appropriate to maximum power level and 1% failed fuel fraction as discussed in Section 11.1.

12.3-29 REVISION 9 - DECEMBER 2002

B/B-UFSAR Shielding thickness are designed with consideration given to all sources in the area including localized hot spots or penetrations. Design parameters are listed in Subsection 12.3.2.1.4. Byron/Braidwood's shielding design is pictured in Drawings M-24-1 through M-24-23. Computer codes used in shielding design account for energy spectra and source strengths for each nuclide (including daughter products), material cross sections or attenuation coefficients for each material or element comprising the shield, dose buildup factors, and other relevant parameters.

12.3.2.1.9 Calculational Methods In the design of the primary shield, the one-dimensional transport code ANISN (Reference 1) was used to calculate the transport of neutrons and gammas from the core. It also analyzed the subsequent production of capture-gamma rays in regions external to the core. The CASK code (Reference 2) coupled neutron-gamma ray library of cross sections was utilized with the ANISN code to enable all production and loss mechanisms for both neutrons and gamma rays to be handled in a single calculation.

The parameters used in the ANISN calculation of the primary shield are given in Table 12.3-4. The fixed neutron source spectrum is given in Table 12.3-5.

Dose rates for siting and shielding design of the OSGSF were determined by calculating the direct dose rate using a point-kernel methodology and the skyshine dose rate using Monte Carlo transport methodology. The analyses used measured dose rates obtained at each steam generator region in conjunction with waste samples to identify the dominant gamma-emitting isotopes (see Table 12.2-55).

All other shields are designed for only gamma-ray attenuation by the standard point attenuation kernel (buildup factor, exponential attenuation, and geometry factor), numerically integrated over the volume of the source. The buildup factors and gamma-ray attenuation coefficients were obtained from published data (References 3 and 4). ISOSHLD-III (Reference 5) and QAD (Reference 6) are two point-kernel computer codes used in this design effort for Byron. For Braidwood design effort three point-kernel computer codes (References 5, 6 and 11) were used.

Tanks, demineralizers, filters and evaporators are generally mocked-up as cylinders with source and source densities homogenized and containing the maximum source volume capacity.

Components containing radioactive water, including demineralizers, evaporators and filters are assumed to have a homogenized source density of 1.0 gm/cc. Tanks containing radioactive gases are assumed to contain their sources at the density of air (1.293 x 10-3 gm/cc). Dimensions are obtained from the vendor drawings of the component.

12.3-30 REVISION 14 - DECEMBER 2012

B/B-UFSAR Spent fuel, charcoal filters, activated reactor internals and head, heat exchangers, radwaste drums, and other radioactive components have more complicated source geometries and source material compositions are more diverse. In all cases, the source is homogenized in order to fit into one of the simpler shielding geometry categories.

For example, the shielding of the radioactive drum storage area is mocked-up using the finite slab geometry option of ISOSHLD.

The source composition is based on the composition of a single radwaste drum (Table 12.2-44) with a reduction factor used for 12.3-30a REVISION 7 - DECEMBER 1998

B/B-UFSAR the packing fraction encountered when cylindrical drums are placed adjacent to each other. Drum storage areas are assumed to be filled to maximum capacity. Sources used for the intermediate activity storage areas are the spent resin decayed for 90 days shown in Table 12.2-43 adjusted to a radionuclide content equivalent to 4.5 ft3 of resin per drum. Sources for the low activity storage areas are based on radwaste evaporator concentrates (Table 12.2-39). Again an adjustment is made to 4.5 ft3 of evaporator concentrate per drum. The resultant shielding of the intermediate and low activity drum storage areas is shown on Drawing M-24-19 and the design basis is given in Table 12.3-2.

Scattered radiation from labyrinths and penetrations is analyzed by the point-kernel single-scatter computer code GGG (Reference

7) or by the Monte Carlo code OGRE (Reference 8). As mentioned in Subsection 12.1.2.2.1, penetrations are located such that direct radiation from the source to the dose point is minimized, and the major contribution to the dose rate is from scattered radiation. If possible, wall penetrations are located above head height, and the use of wall and floor penetrations which run between radioactive areas and unlimited-access areas is minimal.

12.3.2.2 Specific Shielding Design Criteria For purposes of design and operational control, it is necessary and convenient to classify areas (or zones) at the station according to expected personnel access and occupancy requirements. Areas of the station are assigned a design dose rate based on maintaining personnel exposures below 10 CFR 20 limits. Shielding is then designed in conjunction with appropriate radiological access control patterns to assure that area dose rates do not exceed area design dose rates.

Zone classification and dose rate categories for Byron/Braidwood are summarized in Table 12.3-1. Design dose rates for areas surrounding specific equipment and components are set forth in Table 12.3-2.

The shielding design-basis geometries of most major potentially radioactive components are given in Table 12.3-6. The calculations were performed using the computer codes, geometries, and compositions shown. But several considerations in the interpretation of the table need further explanation.

a. The dimensions shown are in many cases, approximate.

However, they have been chosen such that the conservatism of the calculation is not compromised.

The numbers shown are representative shielding design basis only and should not be used for any other purpose.

b. Source compositions are homogenized and tanks are assumed filled to the maximum level to represent the worst case. For some of the heat exchangers and 12.3-31 REVISION 9 - DECEMBER 2002

B/B-UFSAR steam generators the cooling coils were included for self-shielding. For others, cooling coils are ignored for added conservatism.

c. The shell thicknesses of components were in general included in the models for completeness. However, the effect of a fraction of an inch of iron on the gamma radiation considered was found to be insignificant and not included in the table.

d. Where the calculation of ceiling and/or floor shielding thicknesses was necessary, an axial as well as radial case was set up. The table shows representative dimensions for calculation in either direction.

e. The model shown in the table represents one piece of equipment. Where two or more components are in close proximity, dose rates are multiplied by a factor greater than 1 to account for multiple sources. This correction factor is generally equal to the number of components for components in the same cubicle.

f. Pumps are modeled as a pipe which is the same size as the largest pipe attached to the pump. The length of the pipe is determined by the length of the cubicle which houses it.

g. Pipes for pipe tunnel shielding are assumed to contain the same worst case sources as the outlet of the component they are connected to. Multiple pipes are assumed to be carrying their fluid simultaneously.

h. Dose rate detectors are placed on the outside surface of the wall and the dose rate is calculated for at least three thicknesses in each direction in order to obtain a graph of dose rate vs. thickness. This graph is used for choosing a final design shielding thickness using the design-basis dose rates of Table 12.3-2.

i. The source is assumed to be located near the inside surface of the shielding wall to account for associated piping which may run along the wall.

j. The resultant dose rate from all radioactive components in a particular area is considered in the choice of the shielding thickness to meet the design dose rate given on Table 12.3-2. For example, two components which are adjacent to a general access area (1 mr/hr) contribute less than 0.5 mr/hr each.

Expected peak external dose rates throughout the station, covering the two modes of normal plant operation described in 12.3-32

B/B-UFSAR Subsection 12.3.2.1.2, are illustrated on the radiation mapping drawings, Figures 12.3-27 through 12.3-70. The dose rate categories used are given in Table 12.3-1, and each category is mapped on the drawings with a distinct graphic art screen.

The main control room and associated areas under accident conditions are included as a special region on the radiation mapping drawings.

12.3.2.3 Shield Wall Penetrations and Streaming Ratios Penetrations in shield walls for pipes, HVAC ducts, and openings are located and designed to minimize radiation levels to personnel. Location and orientation of penetrations is selected to avoid streaming to areas most likely to be occupied by operating and maintenance personnel.

Compensatory shielding is used where necessary to reduce radiation streaming due to penetrations and localized shield deficiencies (expected hot spots).

Techniques which are used include increased wall thickness, provision for labyrinths or shadow shielding, provision for bends or directing the streaming path away from accessible areas, use of higher density materials such as lead, steel, or lead wool, etc.

Streaming along edges of access hatches, plugs, doors, etc., is minimized by the use of stepped off-sets.

Dose rates from radiation streaming are limited to a peak value at the penetration (i.e., as close as possible to the penetration on the low radiation side of the shield) of:

a. five times the design dose rate for uncontrolled access areas,

b. five times the design dose rate for penetrations located from 0 to 10 feet above the floors in controlled access areas which have design dose rates 10 mrem/hr, and

c. ten times the design dose rate for penetrations located more than 10 feet above the floor in controlled access areas.

For uncontrolled access areas having design dose rates of greater than 10 mrem/hr, specific streaming ratios for penetrations less than 10 feet above the floor are area dependent and may be more or less restrictive than those for controlled access areas of 10 mrem/hr.

12.3-33 REVISION 7 - DECEMBER 1998

B/B-UFSAR The general dose rate which includes radiation streaming, averaged over accessible locations in the protected area, satisfies the design dose rate for the designated area.

Each penetration through a shield wall provides a streaming path for radiation which reduces the shielding effectiveness of the wall, except when the average density of a penetration with a small void content is greater than the average density of the shielding material being penetrated. The magnitude of the reduced effectiveness depends on geometry, material composition, and source characteristics.

In order to minimize the hazard of streaming and to maximum personnel protection, the guidelines listed below are followed in designing and locating shield wall penetrations.

a. Unnecessary penetrations are avoided. A service run or duct is not routed through a shielded cubicle unless that service is provided for equipment within the cubicle.

b. Penetrations are located as far away from radiation sources (e.g., the vessels or piping containing radioactive material) as is practicable.

c. Wherever it is practicable to do so, the penetration is located (1) near where two or three shield walls join, for example, near the upper corners of a room (so that the penetration is far away from radiation sources), and (2) near beams and columns which may serve as extra shielding to at least one side of the penetration (e.g., when beam is between source and penetration).

d. The penetration is located as high above the floor as is practicable and not less than 8 feet if possible.

e. The penetration penetrates through the thinnest of shield walls when a choice exists.

f. The diameter of the penetration is chosen as small as practicable. For electrical penetrations, use of sleeves or conduit having larger than 6-inch nominal diameter is avoided.

g. HVAC ducting avoids penetrating shield walls where practicable. HVAC ducts are routed through the labyrinthine entrances above the doorways of shielded cubicle were feasible. Cases exist, however, where shield wall penetration is necessary. In these cases the proper shielding option(s) to be taken are determined on an individual basis.

12.3-34

B/B-UFSAR

h. If electrical pipe or conduit is routed near the entrance to a radiation source cubicle, advantage is taken where practicable of the HVAC penetrations above the doorway and the conduits are run next to the HVAC control dampers and along the inside walls of the labyrinth and room. (In this case, no shield walls are penetrated.)

i. Where practicable, all pipe and conduit penetrations are grouted.

j. Offset penetrations are used when large lines or ducts penetrate shielding walls of cubicles which contain high levels or radiation, i.e., shield walls greater than or equal to 3-foot thick. HVAC ducts and openings are the most common penetrations that incorporate offsets, but in general, offsets are not used unless no other method will work.

12.3.3 Ventilation Requirements The protective features for the ventilation systems are discussed in detail in Sections 9.4, 11.3, and 11.5.

Drawing M-24-3 (top center) depicts a typical physical layout of the filter systems utilized in the various plant ventilation systems.

Subsection 6.5.1 addresses the operation and design of the engineered safety feature filter systems.

Specific ventilation system designs are discussed in the following subsections:

a. control room HVAC system - Subsection 9.4.1;

b. radwaste building vent system - Subsection 9.4.3.1;

c. laboratory HVAC system - Subsection 9.4.3.2; and

d. auxiliary building HVAC system - Subsection 9.4.5.1.

12.3.3.1 Station Ventilation The design of the station ventilation systems protects plant operating and maintenance personnel and the general public from exposure to radiation from airborne radioactive sources. This requirement applies to all operating conditions, including refueling, maintenance, and anticipated operational occurrences.

For areas other than the control room, the design philosophy is to prevent radioactive contamination of inlet air by preventing release of radioactive contamination to the outside air, instead 12.3-35 REVISION 9 - DECEMBER 2002

B/B-UFSAR of filtering inlet air. Exhaust air from all potentially contaminated areas shall be filtered to meet this philosophy.

Within the station, airflow is normally directed from lesser potential contamination areas to greater potential contamination areas. Areas of greater potential contamination are maintained at a more negative pressure than areas of lesser potential contamination (e.g., general access areas).

The design of the ventilation system for the control room complex is such that, following postulated design-basis accidents, radiation doses to main control room personnel for the duration of the accident will be within the limits set forth in 10 CFR 50, Appendix A, Criterion 19. Radiation protection for the control room consists of adequate air recirculation rates and systems for controlling iodine and particulates in addition to shielding.

Shielding of the main control room is discussed in Subsection 6.4.2.5.

The station ventilation systems are designed so that exhaust air from potentially contaminated areas can be routed through appropriate filters prior to discharge through the ventilation stacks. Stack releases shall be within acceptable limits such that they do not cause offsite doses to exceed the limits set forth in 10 CFR 100 for accidents analyzed using TID-14844, or 10 CFR 50.67 for accidents analyzed using alternative source term methodology, or the limits set forth in 10 CFR 50 Appendix I for normal operating conditions.

Radiation protection considerations for waste filters (which include HVAC filters) are discussed in Subsection 12.3.1.7.

12.3.3.2 Design Criteria To meet the design objectives, the following radiological safety design guidelines were utilized:

a. The system is designed to maintain air flows from clean areas to potentially contaminated areas and from areas of potentially lower level contamination to areas of potentially higher level contamination (prior to exhaust).

b. The system is designed to ensure that negative pressure differential with respect to surrounding areas is maintained inside potentially contaminated cubicles. Control dampers and seals are provided to assure the airflow patterns can be properly maintained.

c. Fume hoods are utilized in the laboratories to facilitate safe processing of radioactive samples by directing contaminants away from the breathing zone to the filtering and ventilation system.

12.3-36 REVISION 12 DECEMBER 2008

B/B-UFSAR

d. Equipment decontamination facilities are ventilated to assure control of released contamination and prevent personnel exposure and the spread of contamination.

e. Exhaust air is routed through HEPA filters or a combination of HEPA and charcoal filters where necessary before release to the atmosphere to reduce onsite and offsite radioactivity levels.

f. Air is supplied to each principal building via separate supply intakes and duct systems.

g. The fresh air supply to the control room is designed to be operable during loss of offsite power. The air is filtered and can be passed through charcoal adsorbers to prevent contamination of the control room by smoke or excessive radioactivity.

h. Transient airborne contamination may result due to maintenance. Special procedures, such as: portable air handling units, and the use of plastic tents is instituted to minimize the contamination on a case by case basis.

i. All exhaust ventilation systems designed to handle potentially contaminated air in the plant are of similar design. A typical filtration system is equipped with a demister and/or prefilter, a heater for humidity control, a set of prefilters, and a set of HEPA filters. Filter systems designed to remove radioiodine are equipped with a charcoal filter bank and an additional set of HEPA filters to collect charcoal fines emerging from the charcoal filters.

Dampers are provided before and after the filter train to isolate the train during filter changes.

j. All filter systems in which radioactive materials could accumulate to produce significant radiation fields external to the ductwork are appropriately located and shielded to minimize exposure to personnel and equipment.

k. Filters in all systems are changed based upon the airflow and the pressure drop across the filter bank. In the case of the prefilters, a pressure drop of 1 inch of water equivalent across the bank is cause for changeout. HEPA filters are changed when the pressure drop across them reaches 2 inches of water equivalent. Charcoal adsorbers are changed based on the residual adsorption capacity of the bed as measured by test samples or canisters removed and analyzed at intervals.

12.3-37

B/B-UFSAR

l. While the majority of the activity in the filter train is removed by simply removing the contaminated filters, further decontamination of the internal structure is facilitated by the proximity of electrical outlets for operation of decontamination equipment, and water supply for washdown of the interior, if necessary. Drains are provided on the filter housing for removal of contaminated water.

These guides are incorporated and fully described in Section 9.4.

12.3-38 REVISION 1 - DECEMBER 1989

BRYON-UFSAR 12.3.3.3 Cubicles Requiring Charcoal Air Filtration Cubicles which contain the following systems or components shall have provisions to exhaust the ventilation air through charcoal filters.

a. post-LOCA recirculation systems;

b. waste filters and demineralizers (see Subsection 12.3.1.7);

c. evaporators for radwaste or recycle; and

d. items with significant concentration of I-131 (more than 0.1 times the I-131 concentration in the reactor coolant, or more than 0.25 Ci/cc, whichever is more limiting).

In general, cubicles containing static tanks and heat exchangers need not have ventilation air passed through charcoal filters since the leakage from such components on cubicle floor is not assumed to have a I-131 concentration exceeding 0.1 times the I-131 concentration in the primary reactor coolant. (The venting of radwaste tanks is through charcoal filters, however, as discussed in Subsection 12.3.1.5).

The following is a list of cubicles which have provisions to pass ventilation air through charcoal filters; leaking equipment in these cubicles could produce levels of airborne I-131 which are one-tenth the levels produced due to leaks in primary coolant equipment.

a. radwaste evaporator cubicles,

b. recycle evaporator cubicles,

c. demineralizer cubicles, and valve aisles

d. primary sample room (local filtration),

e. RHR heat exchanger cubicles,

f. letdown heat exchanger valve aisles, 12.3-38a REVISION 10 - DECEMBER 2004

BRYON-UFSAR

g. centrifugal charging pump cubicles,

h. positive displacement charging pump cubicles,

i. safety injection pump cubicles,

j. auxiliary building equipment drain pump cubicle,

k. waste gas compressor cubicle,

l. gas analyzer cubicle,

m. recycle evaporator feed pump cubicles, and valve aisles

n. gas decay tank cubicles, valve aisles, and pipe tunnel,

o. RHR pump cubicles,

p. containment spray pump cubicles,

q. volume control tank valve aisles,

r. surface condenser rooms,

s. fuel handling building,

t. volume reduction equipment cubicles,

u. radwaste and blowdown mixed bed demineralizer valve aisle, operating area, and cubicles,

v. filter valve aisle, operating area, pipe tunnel, associated filter cubicles, and main area,

w. clothes change and shower room

x. collection drain sump rooms,

y. pipe tunnels, z spray additive tank room and pipe penetration area, aa. CASP room, bb. recycle holdup tank pipe tunnel and tank room, cc. floor drain sump rooms, dd. auxiliary steam pipe tunnels, 12.3-39 REVISION 1 - DECEMBER 1989

BRYON-UFSAR ee. spent resin and concentrates pump room, ff. surface condenser rooms, gg. letdown reheat heat exchanger valve operating area, and hh. HRSS lab area and tank and pump room.

12.3-40 REVISION 1 - DECEMBER 1989

BRAIDWOOD-UFSAR 12.3.3.3 Cubicles Requiring Charcoal Air Filtration Cubicles which contain the following systems or components shall have provisions to exhaust the ventilation air through charcoal filters.

a. post-LOCA recirculation systems;

b. waste filters and demineralizers (see Subsection 12.3.1.7);

c. evaporators for radwaste or recycle; and

d. items with significant concentration of I-131 (more than 0.1 times the I-131 concentration in the reactor coolant, or more than 0.25 Ci/cc, whichever is more limiting).

In general, cubicles containing static tanks and heat exchangers need not have ventilation air passed through charcoal filters since the leakage from such components on cubicle floor is not assumed to have a I-131 concentration exceeding 0.1 times the I-131 concentration in the primary reactor coolant. (The venting of radwaste tanks is through charcoal filters, however, as discussed in Subsection 12.3.1.5).

The following is a list of cubicles which have provisions to pass ventilation air through charcoal filters; leaking equipment in these cubicles could produce levels of airborne I-131 which are one-tenth the levels produced due to leaks in primary coolant equipment.

a. radwaste evaporator cubicles,

b. recycle evaporator cubicles,

c. demineralizer cubicles, and valve aisles

d. primary sample room (local filtration),

e. RHR heat exchanger cubicles,

f. letdown heat exchanger valve aisles,

g. centrifugal charging pump cubicles,

h. positive displacement charging pump cubicles,

i. safety injection pump cubicles,

j. auxiliary building equipment drain pump cubicle,

k. waste gas compressor cubicle, 12.3-40a REVISION 10 - DECEMBER 2004

BRAIDWOOD-UFSAR

l. gas analyzer cubicle,

m. recycle evaporator feed pump cubic aisles

n. gas decay tank cubicles, valve aisl tunnel,

o. RHR pump cubicles,

p. containment spray pump cubicles,

q. volume control tank valve aisles,

r. surface condenser rooms,

s. fuel handling building,

t. volume reduction equipment cubicles,

u. radwaste and blowdown mixed bed demineralizer valve aisle, operating area, and cubicles,

v. filter valve aisle, operating area, pipe tunnel, associated filter cubicles, and main area,

w. mask cleaning room

x. pipe tunnels,

y. spray additive tank room and pipe penetration area,

z. CASP room, aa. recycle holdup tank pipe tunnel and tank room, bb. floor drain sump rooms, cc. auxiliary steam pipe tunnels, dd. spent resin and concentrates pump room, ee. surface condenser rooms, ff. letdown reheat heat exchanger valve operating area, gg. HRSS lab area and tank and pump room, hh. Unit 2 collection drain sump room/hot machine shop, and ii. Unit 1 collection drain sump room.

12.3-40b REVISION 7 - DECEMBER 1998

B/B-UFSAR 12.3.3.4 Ventilation Design Features The ventilation system parameters for radiologically significant areas in the auxiliary building are provided in Table 12.2-45.

12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation Two fixed systems are provided to monitor radiation/radioactivity levels within the plant. These are:

a. the area radiation monitoring system (ARMS), and

b. the continuous airborne monitoring system (CAMS).

Portable CAMS, grab sampling capability, and automatic samplers are also provided to supplement the fixed monitoring systems.

The fixed ARMS is provided to continuously measure, indicate, and trend the levels of radiation in general access and operational areas. Radiation alarms are activated when predetermined levels are exceeded. The objective is to keep operating personnel informed of the radiation levels in the selected areas and thus assist in avoiding unnecessary or inadvertent exposure.

The fixed CAMS is provided to measure, indicate, and trend the levels of airborne radioactivity in the air exhausted from cubicles or branch HVAC exhaust ducts. The objective is to warn operators that airborne activity may be present in the area or cubicle serviced by the monitored exhaust system, and thereby assist in avoiding unnecessary or inadvertent exposure. CAMS also provides a means for identifying trends in air concentration levels and the source of the activity. Each fixed CAM activates visual and audible control room alarms when predetermined levels are exceeded. The fixed monitors are also used to assist in the monitoring and control of effluents as described in Section 11.5. Portable CAMS are generally used to monitor the ambient air in normally occupied areas. They may be used in conjunction with the fixed CAMS to locate the source of the airborne activity.

12.3.4.1 Area Radiation Monitoring Instrumentation The area radiation monitoring system (ARMS) is provided to fulfill the following specific radiological safety objectives:

12.3-40c REVISION 1 - DECEMBER 1989

B/B-UFSAR

a. To provide operating personnel in the main control room with an indication and record of radiation levels at selected locations within the various plant buildings (e.g., to warn of excessive radiation levels in areas where nuclear fuel is stored or handled).

b. To contribute radiation information to the control room so that correct decisions may be made with respect to deployment of personnel in the event of a radiation incident.

c. To assist in the detection of unauthorized or inadvertent movement of radioactive material in the plant including the radwaste area.

d. To supplement other systems in detecting abnormal migrations from and radioactive material in the process streams.

e. To provide local alarms at key points where a substantial change in radiation levels might be of immediate importance to personnel frequenting the area. Area monitors in high noise areas feature visual as well as audible alarms.

f. To assist in maintaining in-plant personnel exposure as low as reasonably achievable.

To implement these objectives area radiation detectors are provided throughout the plant at locations indicated in Table 12.3-3 and shown on the radiation shielding design drawings in Section 12.3 (see Drawings M-24-1 through M-24-23).

ARM's are installed in the vicinity of the fuel pool and on the fuel handling building overhead crane and in containment to sense abnormal or accident conditions as indicated in Table 12.3-3.

The ranges and initial setpoints are also given in Table 12.3-3.

The general requirements for the ARM's are as follows:

Energy Response Gamma energy response of the detectors extends from 0.02 to 3 MeV. The energy dependence is within 20%.

Channel Accuracy The overall channel accuracy within the environmental limitations of the system is 20% or better of reading (digital output).

12.3-41 REVISION 9 - DECEMBER 2002

B/B-UFSAR Precision The reproducibility of each channel for any given measurement over its stated range is 10% or better at the 95% confidence level.

Power Supply Area radiation monitors receive power from 120-Vac buses. The audio and visual alarms receive power from the same 120-Vac buses. Nuclear safety-related area radiation monitors receive power from ESF buses.

Calibration Area radiation monitor calibration frequency is established based on safety significance of the application and equipment historical performance. Area radiation detectors have the capability of being calibrated for dose rate in the calibration facility by exposing the detectors to the radiation field from an isotope of known activity. Cabling is provided from the calibration facility to the control room permitting readout in the control room. The intensity of the calibration field can be varied, thereby allowing a multipoint calibration.

Location Location of area radiation detectors is provided in Table 12.3-3.

Each area radiation detector is connected to the main control room central processor by microprocessors (monitors) which control the detectors, process and store its data. The central processing console for each unit includes a video display unit.

Each monitor maintains trend files that can be accessed through the central processing console.

Dedicated readout modules and recorders are provided for those nuclear safety-related area radiation monitors only, whose application in the plant design requires a safety-related operator interface and/or data collection capability. All monitors are designed to fail in the safe (Alarm) mode.

Conformance to Applicable Regulations The ARMS conforms to Sections 4.2 and 5.3.4.1 of ANSI N13.2-1969.

Qualified personnel have been used in the engineering phase and will be used during operation to assure that radiation exposures to plant personnel will be ALARA. Regulatory guidance concerning effluents and ANSI N13.1-1969, do not directly apply to the ARMS.

12.3.4.2 Continuous Airborne Monitoring Instrumentation The continuous airborne monitoring system (CAMS) is provided for monitoring in-plant airborne radioactivity levels. The specific radiological safety objectives are the same as Subsection 12.3.4.1. Continuous air monitors (CAMs) are discussed in Section 12.3-42 REVISION 17 - DECEMBER 2018

B/B-UFSAR 11.5 and identified in Table 11.5-1, including location, range, sensitivity, and alarm setpoints. Monitor locations are shown and identified in Drawings M-24-1 through M-24-22. Probe locations are shown on the HVAC system drawings in Section 9.4.

The fixed continuous airborne monitors (CAMs) are provided to monitor for airborne radioactivity in compartments which may be occupied and may contain airborne radioactivity. Since there are too many rooms or cubicles to monitor independently, a limited number of CAMs are provided to continuously monitor the air from selected branch exhaust ducts of the HVAC system.

The exhaust from a single room may be diluted by the exhaust from other rooms before the air gets to the monitoring point.

Therefore, the monitor must be sensitive enough to respond to the diluted activity. The maximum possible dilution factor for any cubicle is:

F cubicle (cfm)

DF F duct (cfm)

Using the detectability factor for MPCa (DMPCa) given in Table 12.3-9, an expression can be written for the time, T, it takes to detect the presence of MPCa levels in the exhaust ducts (The term MPC as used in this section refers to a 10CFR20 limit in effect prior to January 1, 1994):

T = 1/(DMPCa

DF), (hrs)

where, T = time to detect particulate and iodine MPCa, (hr),

DMPCa = detectability factor for MPCa (see Table 12.3-9),

DF = dilution factor for 10 MPC-HR detectability, Fcubicle = flow in exhaust from cubicle, (cfm), and Fduct = flow in branch duct where monitor is located, (cfm).

12.3-43 REVISION 9 - DECEMBER 2002

B/B-UFSAR Table 12.3-9 shows the sensitivity of the particulate, iodine, and noble gas channels for the isotopes of greatest interest.

These sensitivities were compared to maximum permissible concentrations in air (MPCa) of the most restrictive particulate and iodine radionuclides in the areas and cubicles of lowest ventilation flow rate. The criterion used was that airborne radioactivity from the areas described above and having an activity concentration of one MPCa would be detected within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . Exhaust flow rates from cubicles and in branch ducts were examined to determine dilution factors for this assessment. The exhaust flow rates for the monitored branch ducts and the individual room exhaust flow rates are given in the drawings cited 12.3-43a REVISION 9 - DECEMBER 2002

B/B-UFSAR in Section 9.4. The location of the radiation monitors are also shown on these drawings. An investigation using the above data indicates that the system is capable of detecting 10 MPCa-hrs of airborne particulate and iodine radioactivity in the rooms, cubicles, and areas discussed above which may be occupied and may contain airborne radioactivity.

The general requirements for the CAMS are as indicated in the following.

Energy Response of Channels Gamma energy response of the detector channels used for gamma monitoring extends from 0.08 to 3 MeV. The energy dependence is within 20%. Beta detector channels are capable of detecting minimum of 0.07 MeV beta (e.g., 7 mg/cm2 aluminum window).

Channel Accuracy The accuracy of each channel is within 20% or better of the reading.

Precision The precision is 10% or better at the 95% confidence level.

Particulate Filters Filters have an efficiency of 99% or better for 0.3 micron particles.

Iodine Collector Iodine collectors consist of activated, impregnated charcoal cartridges in metal canisters. Prefilters are installed upstream of cartridges to remove particulates.

Representative Sampling Design Sampling systems are designed to assure representative sampling for off-line CAM. Isokinetic sampling nozzles are used for extraction of gaseous samples from gaseous streams. The in-duct isokinetic probes comply to the standard set forth in ANSI N13.1-1969. Sample piping is designed to avoid sharp bends and stagnant zones. Off-line detector assemblies are designed with temperature, pressure, and flow regulators as required for instrumentation. All off-line monitors are capable of being purged with air.

Power Supply Electric power is provided to CAMs from permanent supplies.

12.3-44 REVISION 9 - DECEMBER 2002

B/B-UFSAR Alarms The CAMs are provided with two adjustable alarm setpoints (alert and high alarms). There is also an instrument failure alarm.

Each of the above indicates in the control room and has a relay contact output at the microprocessor cabinet.

Periodic Testing All CAMs are capable of being checked, tested, and calibrated periodically to verify proper operation. Check sources, test signals, and calibration sources are provided as applicable.

It is possible to periodically test those CAM's, which are related to nuclear safety in accordance with criteria for periodic testing of protection system actuation functions and IEEE 338-1971. Such testability means the ability to duplicate required functions as closely as possible (e.g., during reactor operation) without impairing plant operation. The air sample calibration programs comply with the guidance contained in ANSI guide IEEE N232C Section 4.5.

All trip circuits are capable of convenient operational verification by means of test signals or through the use of portable sources.

Radionuclide standards of two or more different source strengths are provided. Gaseous detectors requiring in-place radiogas calibration are provided with necessary isolation valves.

Recirculation design is employed to minimize gas usage.

The shield assembly is designed to allow quick and simple purging, decontamination and removal of sample canister, and replacement with standard canister.

12.3.5 References

1. W. W. Engle, Jr., "A Users Manual for ANISN, A One-Dimensional Discrete-Ordinates Transport Code with Anisotropic Scattering," K-1963, Union Carbide Corporation, Nuclear Division, March 30, 1967.

2. RSIC Data Library, "DIC-23/CASK 40 - Group Coupled Neutron and Gamma-Ray Cross Section Data," Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee, December 1972.

3. J. H. Hubbel, "Photon Cross Sections, Attenuation Coefficients, and Energy Absorption Coefficients from 10 keV to 10 GeV," NSRDS-NBS29, August 1969.

4. S. Buscagline and R. Manzini, "Buildup Factors: Coefficients of the J. J. Taylor Equation," ORNL-tw-80, February 1964.

12.3-45 REVISION 14 - DECEMBER 2012

B/B-UFSAR

5. R. L. Angle, J. Greenborg, and M. M. Hendrickson, "ISOSHLD -

A Computer Code for General-Purpose Isotope Shielding Analysis,"

BNWL-236, Pacific Northwest Laboratory, Richland, Washington, June 1966, Supplement 1, March 1977, Supplement 2, April 1969.

6. R. E. Malenfant, "QAD: A Series of Point-Kernel General-Purpose Shielding Programs," LA-3573, Los Alamos Scientific Laboratory, April 5, 1967.

7. R. E. Malenfant, "G3: A General-Purpose Gamma-Ray Scattering Program," LA-5176, Los Alamos Scientific Laboratory, June 1973.

8. D. K. Trubey and M. B. Emmett, "OGRE-G, A General-Purpose Monte Carlo Gamma-Ray Transport Code," ORNL-TM-1212, Oak Ridge National Laboratory, 1966.

9. WCAP-8872, "Design, Inspection, Operation and Maintenance Aspects of the Westinghouse NSSS to Maintain Occupational Radiation Exposures As Low As Reasonably Achievable," April 1977.

10. W. J. Manion and T. S. LaGuardia, "An Engineering Evaluation of Nuclear Power Reactor Decomissioning Alternatives," Atomic Industrial Forum, Inc., Document No. AIF/NESP-009, November 1976.

11. Braidwood only - MicroShield 8.01, Grove Software RadiationSoftware.com, 2008, Incorporated in URS qualified software library, April 2009.

12.3-46 REVISION 14 - DECEMBER 2012

B/B-UFSAR TABLE 12.3-1 CLASSIFICATION OF RADIATION ZONES FOR SHIELD DESIGN AND RADIOLOGICAL ACCESS CONTROL GENERALIZED ZONE DESIGN DOSE RADIOLOGICAL NRC POSTING DESIGNATION RATE* (mrem/hr) TYPICAL REGIONS ACCESS CONTROL REQUIRED I-A 0.2 Plant grounds outside security fencing and Per station procedures None office areas I-B 0.5 Most plant grounds within security fencing Per station procedures None (also OSGSF roof)

I-C 1 Most operating areas and Per station procedures None passageways I-D 2 Assigned as required in Per station procedures None design II-A 4 Assigned as required in Per station procedures None design II-B 10 Assigned as required in Per station procedures Radiation area design

For a given operating mode, the design dose rate is the maximum dose rate expected after 10 years of plant operation in a given region outside highly localized radiation streaming paths.

12.3-47 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.3-1 (Cont'd)

GENERALIZED ZONE DESIGN DOSE RADIOLOGICAL NRC POSTING DESIGNATION RATE* (mrem/hr) TYPICAL REGIONS ACCESS CONTROL REQUIRED II-C 20 Assigned as required in Per station procedures Radiation area design II-D 100 Assigned as required in Per station procedures Radiation area design III >100 Generally a source region Per station procedures High radiation area IV Not Assigned V 0.2 Main control room Per station procedures None (Normally)

Postaccident** Main control room Per station procedures As required

For a given operating mode, the design dose rate is the maximum dose rate expected after 10 years of plant operation in a given region outside highly localized radiation streaming paths.

- For the initial, 30-day, postaccident period, the design doses to personnel during access and occupancy of the control room are limited to a maximum of 5 rem to the whole body, 30 rem to the thyroid, 30 rem to the bone, and 15 rem to the lung for accidents analyzed using TID-14844. For accidents analyzed using alternative source term methodology, radiation exposure limits are provided in 10 CFR 50.67.

12.3-48 REVISION 12 - DECEMBER 2008

B/B-UFSAR TABLE 12.3-2 SPECIFIC SHIELDING DESIGN CRITERIA RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

I. Reactor containment building I.C.1 Reactor containment 17Z6 Normal operation Area outside 1Z1 0.5 mrem/hr I-B 1A9-1 building building 2A9-1 I.C.2 Reactor containment 18Z10 Normal operation (1) outside per- 5Z2 4 mrem/hr II-A 1A6-1 and equipment sonnel air lock 5Z12 2A6-1 El. 426 ft - 0 in. and equipment hatch El. 401 ft - 0 in. 17Z6 Normal operation (2) emergency 17Z8 1 mrem/hr I-C 1C5-10 personnel 2C5-10 air lock TABLE DEFINITIONS

- This is the design dose rate for the protected area and does not include the contribution of any radioactive components that might be in the area.

- - Hot spot criteria x - Design dose rate is proportional to RBP given in parentheses.

xx - Same as xxx except that the protected area is a radiation area.

xxx - This protected area is a high radiation area due to the presence of radioactive valves and piping. These sources hinder detection of radiation from the shielded source and would cause high personnel exposure. Therefore a RBP during startup testing would only achieve unnecessary personnel exposure.

N/A - Not applicable to start up testing; the zone designation in brackets is the expected level following shutdown or during refueling.

HOT SPOT - The dose rate near penetration in shield walls are permitted to be five times the dose rate specified for the shield wall. For additional information on the hot spot criteria, see Subsection 12.3.2.3.

TML - Too many zones to list.

12.3-49

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP) 1.C.3 Reactor containment 18Z10 Normal operation Area outside 3Z12 2 mrem/hr I-D 1A7-9 and equipment purge penetration containment in 3Z14 10 mrem/hr ** 2A7-9 auxiliary bldg.

on El. 451'-0" I.P. Primary shield I.P.1 Reactor core and 15Z1 Normal operation Outside center- 15Z2 reactor pressure plane of core vessel (1) neutrons outside primary 5 mrem/hr III xxx plus gammas shield (2) thermal 1 x 105 III xxx neutrons neutrons/cm2-sec

(<1.12eV)

(3) epithermal 6.5 x 103 III xxx neutrons neutrons/cm2-sec (1.12 eV<E3.35 keV)

(4) fast - 7.5 x 103 III xxx neutrons neutrons/cm2-sec (E>3.35 keV)

I.P.2 Reactor core and 15Z1 (1) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Outside primary 15Z2 25 mrem/hr [II-D] N/A reactor pressure after shutdown shield vessel (shutdown)

(2) 1 day after Outside primary 10 mrem/hr [II-B] N/A shutdown shield 12.3-50

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

I.M. Missile wall (secondary shield)

I.M.1 Reactor coolant 15Z2 (1) normal Outside missile 15Z3 20 mrem/hr II-D 1,2C3-3 loop, steam 16Z1 operation wall and fan 16Z3 1,2C4-2 generators, and 17Z5 cooler 17Z6 1,2C5-2 containment sump penetrations, 1,2C6-5 (2) 1 day after Outside missile 2 mrem/hr [I-D] N/A shutdown wall I.RP. Refueling cavity I.RP.1 Upper internals 17Z1 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after Outside storage 17Z5 5 mrem/hr [III] N/A shutdown area I.RP.2 Lower internals 17Z1 1 week after Outside storage 17Z5 5 mrem/hr [III] N/A shutdown area I.RP.3 Spent fuel assembly 17Z1 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after (1) outside 16Z1 100 mrem/hr [III] N/A and RCC elements, shutdown refueling canal, reactor and reactor inside missile cavity pool water wall (2) 2 feet above 18Z10 2 mrem/hr [II-A] N/A water level (3) outside fuel 16Z3 1 mrem/hr [I-C] N/A transfer canal, outside missile wall 12.3-51

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

I.B. Reactor containment building general I.B.1 Reactor coolant 15Z2 Normal operation Area outside 15Z3 20 mrem/hr II-D 1C3-2 loop missile wall 2C3-2 I.B.2 Reactor coolant 15Z2 (1) normal Area outside 15Z3 20 mrem/hr II-D 1C3-4 drain, tanks and operation missile wall 2C3-4 pumps (2) during Area outside 2 mrem/hr [II-D] N/A refueling missile wall I.B.2 Containment sump 15Z2 Shutdown Area outside 15Z3 2 mrem/hr [II-D] N/A pumps missile wall I.B.3 Incore instrument 17Z2 (1) normal Area outside 17Z6 20 mrem/hr II-D XXX shaft and storage 15Z2 operation missile wall 15Z3 area (2) during Area outside 2 mrem/hr [I-D] N/A refueling missile wall I.B.4 Regenerative heat 17Z3 (1) normal Area outside 17Z6 20 mrem/hr II-D 1C5-4 exchangers and 17Z4 operation missile wall excess letdown heat exchangers N/A (2) shutdown Area outside 2 mrem/hr [I-D] N/A heat exchanger cubicle 12.3-52

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

I.B.5 Seal table and 17Z2 (1) normal Area outside 17Z6 20 mrem/hr II-D 1C5-3 core detector operation missile wall 100 mrem/hr ** 2C5-3 storage (2) shutdown Area outside 2 mrem/hr [I-D] N/A cubicle I.B.6 Containment char- 18Z10 (1) after Area outside 18Z10 5 mrem/hr II-D xx coal filter system cleanup mode filter housing (2) during Area outside 2 mrem/hr [II-D] N/A refueling filter housing I.B.7 Main steam pipe 16Z5 Normal Area outside 17Z6 20 mrem/hr II-D 1C5-6 chases 16Z6 operation pipe tunnel 2C5-6 II. Fuel handling building II.SF.1 Spent fuel pit 14Z2 Containing 1-2/3 (1) outside spent 14Z4 15 mrem/hr [III] N/A core with 1 week fuel pool wall near decay heat exchangers (2) pipe tunnel 14Z1 15 mrem/hr [III] N/A on sides of 14Z9 spent fuel pool (3) inside fuel 14Z10 50 mrem/hr [III] N/A transfer canal (4) 2 feet above 13Z8 4 mrem/hr [II-A] N/A water level 12.3-53

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

II.SF.2 Spent fuel trans- 14Z10 Spent fuel (1) outside of 12Z5 15 mrem/hr [III] N/A fer canal assembly with canal 12Z10 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months decay (2) 2 feet above 13Z9 2 mrem/hr [I-D] N/A water level II.SF.3 Spent fuel pit heat 14Z4 Normal operation Area outside heat 14Z6 2 mrem/hr I-D F5-2 exchangers exchanger cubicle F5-3 II.SF.4 Spent fuel pit pumps 14Z4 Normal operation Area outside pump 14Z6 2 mrem/hr I-D F5-3 and sump cubicle II.SF.5 Spent fuel pit 14Z4 Normal operation Area outside pump 14Z6 2 mrem/hr I-D Ft-3 skimmer pump cubicle II.SF.6 Fuel handling 14Z6 Normal operation Area outside 1Z1 0.5 mrem/hr I-B F5-5A building building F5-5B III. Auxiliary building III.1 Elevation 330 ft. - 0 in.

III.1.1 Auxiliary building 10Z1 Filled with con- Area outside 10Z4 2 mrem/hr I-D 1A1-1 sumps 10Z5 taminated water cubicles 10Z8 2A1-1 III.1.2 Auxiliary building 10Z3 Pumping contami- Outside pump 10Z4 2 mrem/hr I-D 1A1-2 equip. drain pumps 10Z7 nated water rooms 10Z8 2A1-2 12.3-54

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.2 Elevation 346 ft - 0 in.

III.2.1 Recycle holdup 9Z14 Filled with Area outside 9Z1 2 mrem/hr I-D OA2-14A tanks 9Z13 contaminated cubicle OA2-14C water III.2.2A Gas decay tanks 9Z6 Filled with (1) area outside 9Z1 2 mrem/hr I-D OA2-17 fission product tank cubicles gases (2) valve aisle 9Z2 15 mrem/hr III xxx III.2.2B Waste gas valve 9Z2 Normal operation (1) valve 9Z1 2 mrem/hr I-D OP2-34 thru aisle operating area 10 mrem/hr ** OP2-38 (2) area outside 9Z1 2 mrem/hr I-D OA2-15 cubicle OA2-16 III.2.3 Auxiliary building 9Z18 Filled with con- Area outside pump 9Z1 2 mrem/hr I-D 1A2-12 collection sump 9Z35 taminated water and sump area 2A2-12 pumps III.2.4 Auxiliary building 9Z19 Filled with con- Area outside 9Z1 2 mrem/hr 1-D 1A2-10 equipment drain 9Z20 taminated water cubicles 2A2-10 tanks III.2.5 Containment spray 9Z7 (1) Normal Area outside 9Z27 15 mrem/hr III xxx pumps 9Z8 operation pump cubicle 9Z28 (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Area outside 100 mrem/hr - -

after LOCA pump cubicle 12.3-55

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.2.6 Moderating heat ex- 9Z21 Normal (1) area out- 9Z1 2 mrem/hr I-D 1A2-3,4,5,6,9 changers, letdown 9Z25 operation side cubicle 2A2-3,4,5,6,9 chiller and letdown 9Z33 reheat heat ex- 9Z22 (2) valve opera- 9Z37 4 mrem/hr II-A 1A2-7,8 changers 9Z26 ting area 9Z36 2A2-7,8 9Z34 (3) valve aisle 9Z23 15 mrem/hr III xxx 9Z24 III.2.7 Recycle evaporator 9Z15 Pumping con- (1) outside pump 9Z1 2 mrem/hr I-D OA2-14A feed pumps 9Z16 taminated water cubicle OA2-14C (2) valve opera- 9Z1 2 mrem/hr I-D OA2-14B ting area (3) valve aisle 9Z17 15 mrem/hr II-D xx III.2.8 Pipe tunnel for 11Z7 Contaminated Area outside 9Z1 2 mrem/hr I-D OA2-13C recycle evaporator water, normal pipe tunnel operation III.2.9 Recycle evaporator 9Z9 Operation of Outside evaporator 9Z1 2 mrem/hr I-D OA2-13A&B packages 9Z11 evaporators package cubicles 10 mrem ** OP2-31&32 III.2.10 Waste gas pipe 11Z7 (1) Pipes con- Area outside pipe 8Z1 2 mrem/hr I-D OA3-18A tunnel taining contami- tunnel nated gas 12.3-56

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.2.11 RHR pumps 9Z27 (1) operation at Entrance to pump 9Z7 15 mrem/hr [III] N/A 9Z28 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after cubicle 9Z8 shutdown (2) gap release Area outside 9Z7 0.5 rem - -

accident, any 8 pump cubicle 9Z8 after a gap hour period release accident (3) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Area outside pump 9Z7 100 mrem/hr - -

after LOCA cubicle 9Z8 III.3 Elevation 364 ft 0 in.

III.3.1 Recycle holdup 8Z11 Filled with con- (1) outside tank 8Z1 2 mrem/hr I-D OA3-12A tanks 8Z12 taminated water cubicle OA3-12B (2) valve 8Z1 2 mrem/hr I-D OA3-12C operating area III.3.2 RHR heat exchangers 8Z18 (1) operation 4 Outside heat ex- 8Z1 2 mrem/hr [I-D] N/A 8Z28 hours after changer cubicle 8Z6 shutdown 8Z21 (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Outside heat ex- 100 mrem/hr - -

after LOCA changer cubicle 12.3-57

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

(3) Operation at Outside heat ex- 5 mrem/hr - -

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after a changer cubicle gap release accident III.3.3 Safety injection 12Z6 Pumping sump Outside pump 12Z5 100 mrem/hr - -

12Z9 water, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months cubicle 12Z10 after LOCA III.3.4 Blowdown condensers 8Z6 Primary system Outside condenser 8Z1 2 mrem/hr I-D OA3-14A 8Z10 leakage of 1 gpm cubicle OA3-14B with a total blow-down flow for each unit at 135 gpm III.3.5 Chemical drain 8Z2 Filled with con- Outside tank 8Z1 2 mrem/hr I-D OA3-17 tank taminated water cubicle III.3.6 Chemical drain 8Z3 Pumping con- Outside pump 8Z1 2 mrem/hr I-D OA3-16 pumps taminated water cubicle III.3.7 Auxiliary building 8Z7 Filled with con- Outside tank 8Z1 2 mrem/hr I-D OA3-11B floor drain tanks 8Z8 taminated water cubicle 10 mrem/hr ** OP3-29 III.3.8 Auxiliary building 8Z9 Pumping con- Outside pump 8Z1 2 mrem/hr I-D OA3-13 floor drain pumps taminated water cubicle III.3.9 Vertical pipe 8Z16 Pipes contain- Area outside 8Z1 2 mrem/hr I-D x(OA4-16B) tunnel ing waste gas pipe tunnel 12.3-58

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.3.10 Charging pumps 8Z20 Normal Outside pump 8Z1 2 mrem/hr I-D IA3-5 8Z25 Operation cubicle 2A3-5 Pumping sump Outside pump 8Z1 100 mrem/hr - -

water 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months cubicle after LOCA III.3.11 Chemical/Regener- 8Z4 Demineralizer Outside cubicle 8Z1 2 mrem/hr I-D OA3-16 ation waste drain regenerants (10 mrem/hr) ** OP3-35 tank III.3.11.A Chemical/Regener- 9Z5 Demineralizer Area above 8Z1 2 mrem/hr I-D OA3-18B ation waste drain regenerate cubicle tank removable slab III.3.12 Chemical/Regener- 8Z5 Pump deminera- Outside cubicle 8Z1 2 mrem/hr I-D OA3-15 ation waste drain lizer regenerants pumps III.3.13 Pipe tunnel 11Z4 (1) radioactive Area outside 8Z1 2 mrem/hr I-D OA4-28 El. 375 ft - 6 in. water, normal pipe tunnel operation (2) contaminated Area outside pipe 8Z1 100 mrem/hr - -

piping at 12 tunnel hours after LOCA III.3.14 Pipe tunnels 11Z6 Normal operation Area outside 8Z1 2 mrem/hr I-D OA3-11A El. 374 ft - 6 in. tunnel 12.3-59 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.4 Elevation 383 ft - 0 in.

III.4.1 Anion and cation 7Z10 Contaminated Outside deminera- 7Z1 2 mrem/hr I-D OP4-25 demineralizers demineralizer lizer cubicle 10 mrem/hr **

III.4.2 Blowdown mixed bed 7Z10 Contaminated (1) valve aisle 7Z12 4 mrem/hr III xxx demineralizers demineralizer (2) valve opera- 7Z20 4 mrem/hr II-A OA4-5 ting area III.4.3 Radwaste mixed bed 7Z11 Contaminated (1) valve aisle 7Z12 4 mrem/hr III xxx demineralizer demineralizer (2) valve opera- 7Z20 4 mrem/hr II-A x(OA4-5) ting area III.4.4 Anion filters 10Z42 Contaminated (1) area outside 7Z1 2 mrem/hr I-D x(OA4-5) filter filter cubicle (2) valve opera- 7Z20 4 mrem/hr II-A x(OA4-5) ting area (3) pipe tunnel 7Z13 15 mrem/hr III xxx III.4.5 Cation filters 10Z41 Contaminated (1) pipe tunnel 7Z13 15 mrem/hr III xxx filter (2) valve opera- 7Z20 4 mrem/hr II-A x(OA4-5) ting area 12.3-60

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.4.6 Blowdown mixed bed 10Z38 Contaminated (1) pipe tunnel 7Z13 15 mrem/hr III xxx demineralizer after filter filters, and rad- (2) valve opera- 7Z20 4 mrem/hr II-A x(OA4-5) waste filters ting area III.4.7 Recycle evaporator 10Z43 Contaminated (1) pipe tunnel 7Z15 15 mrem/hr III xxx condensate filter filter (2) valve opera- 10Z36 4 mrem/hr II-A x(OA4-7) ting area III.4.8 Recycle evaporator 7Z14 Contaminated (1) area outside 7Z1 2 mrem/hr I-D OA4-6 filter valve aisles water in pipes valve aisles OA4-8 and valves OA4-11B (2) Valve opera- 7Z20 4 mrem/hr II-A x(OA4-5) ting aisle 10Z36 x(OA4-7)

OA4-11A III.4.9 Recycle evaporator 10Z43 Contaminated (1) pipe tunnel 7Z15 15 mrem/hr III xxx feed filters filter (2) valve opera- 10Z36 4 mrem/hr II-A x(OA4-7) ting area III.4.10 Seal water return 10Z43 Contaminated (1) pipe tunnel 7Z15 15 mrem/hr III xxx filters filter (2) valve opera- 10Z36 4 mrem/hr II-A OA4-7 ting area 12.3-61

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.4.11 Seal water injec- 10Z43 Contaminated (1) pipe tunnel 7Z16 15 mrem/hr III xxx tion filters filter (2) valve opera- 10Z36 4 mrem/hr II-A x(OA4-7) ting area III.4.12 Reactor coolant 10Z43 Contaminated (1) pipe tunnel 7Z15 15 mrem/hr III xxx filters filter (2) valve opera- 10Z36 4 mrem/hr II-A x(OA4-7) ting area III.4.13 Vertical HVAC 7Z40 Contaminated Area outside 7Z1 2 mrem/hr I-D OA4-29 Pipe Tunnel water in piping tunnel and airborne in duct III.4.14 Spent fuel pit 10Z43 Contaminated (1) pipe tunnel 7Z16 15 mrem/hr III xxx and skimmer filter filters (2) Outside top 6Z1 2 mrem/hr I-D OA5-23 of filter cubicles III.4.15 Blowdown pre- 10Z40 Contaminated (1) area outside 7Z1 2 mrem/hr I-D OA4-8 filters filter filter cubicle (2) valve opera- 7Z20 4 mrem/hr II-A x(OA4-5) ting area (3) pipe tunnel 7Z15 15 mrem/hr III xxx 12.3-62

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.4.16 Auxiliary building 10Z47 Contaminated (1) pipe tunnel 7Z13 15 mrem/hr III xxx floor drain filter filter (2) valve opera- 7Z20 4 mrem/hr II-A x(OA4-5) ting area III.4.17 Auxiliary building 10Z46 Contaminated (1) pipe tunnel 7Z13 15 mrem/hr III xxx equipment drain filter filter (2) valve opera- 10Z36 4 mrem/hr II-A x(OA4-7) ting area III.4.18 Regeneration waste 10Z45 Contaminated (1) area outside 7Z1 2 mrem/hr I-D x(OA4-10) drain filter filter filter cubicle (2) pipe tunnel 7Z16 15 mrem/hr III xxx III.4.19 Chemical drain 10Z45 Contaminated (1) area outside 7Z1 2 mrem/hr I-D OA4-10 filter filter filter cubicle (2) pipe tunnel 7Z16 15 mrem/hr III xxx III.4.20 Drumming stations 7Z5 Operation of (1) outside drum- 7Z1 2 mrem/hr I-D OA4-14 and pipe tunnel 7Z7 drum processing ming station 11Z14 system (2) maintenance 7Z3 4 mrem/hr II-A OA4-13 aisle 12.3-63

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.4.21 Drumming station 7Z2 (1) Transporting Outside tunnel 7Z43 1 mrem/hr I-C OP4-27 conveyor tunnel of drums filled at 24-wall and 5 mrem/hr **

with contaminated N-wall material (2) Drumming Radwaste and 7Z43 1 mrem/hr I-C OA4-15A station shutdown control OA4-15B operation rooms III.4.22 Letdown heat 7Z27 Reactor coolant (1) outside heat 7Z1 2 mrem/hr I-D IA4-1 exchangers and 7Z30 in tube side of exchanger 2A4-1 seal water heat 7Z34 heat exchangers cubicles 1A4-4 exchangers 7Z37 2A4-4 7Z31 7Z38 (2) valve opera- 7Z29 4 mrem/hr II-A 1A4-3 ting area 7Z36 2A4-3 (3) valve aisle 7Z28 15 mrem/hr III xxx 7Z35 III.4.23 RHR heat ex- 7Z26 (1) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Outside cubicle 7Z1 2 mrem/hr [I-D] N/A changers 7Z33 after shutdown 7Z32 7Z39 (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Outside heat ex- 7Z1 100 mrem/hr - -

after LOCA changers cubicle 12.3-64

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

(3) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after Outside cubicle 7Z1 5 mrem/hr - -

a gap release accident III.4.24 Pipe tunnels 7Z13 Pipes containing (1) area outside 7Z1 2 mrem/hr I-D x(OA4-9) 7Z15 radioactive water, pipe tunnels x(OA4-10)f 7Z16 normal operation (2) valve aisles 7Z12 4 mrem/hr III xxx 7Z14 7Z17 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Area outside 7Z1 100 mrem/hr - -

after LOCA tunnel III.4.25 Vertical waste 7Z41 Normal operation Area outside 7Z43 1 mrem/hr I-C OA4-16B gas tunnel tunnel III.4.26 Pipe tunnel 11Z3 (1) radioactive Area outside 7Z1 2 mrem/hr I-D ORE-AR008 El. 394 ft - 6 in. water pipes, pipe tunnel Area Rad normal operation Monitor (2) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Area outside 7Z1 100 mrem/hr - -

after LOCA pipe tunnel III.4.27 Pipe tunnel 11Z5 Contaminated Area outside 7Z1 2 mrem/hr I-D OP4-27 El. 394 ft - 0 in. water and sludge tunnel 10 mrem/hr **

III.5 Elevation 401 ft - 0 in.

III.5.1 Primary sample 6Z14 Radioactive Outside room 6Z1 2 mrem/hr I-D x(OA5-10A) room samples 12.3-65

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.5.2 Sample heat 6Z18 Cooling radio- (1) primary 6Z14 4 mrem/hr II-A 1A5-11 exchangers active samples sample room (2) outside 6Z1 2 mrem/hr I-D OA5-10A sample room (3) entrance 6Z14 4 mrem/hr II-A OA5-10B to sample cooler III.5.3 Dumbwaiter TML Radioactive Outside shaft 6Z14 4 mrem/hr II-A 2A5-11 samples III.5.4 Thermal regenera- 10Z31 Contaminated Pipe tunnel 6Z26 15 mrem/hr III xxx tion demineralizers demineralizer 6Z27 III.5.5 Recycle evaporators 10Z35 Contaminated (1) area outside 6Z1 2 mrem/hr I-D 1,2A5-5 condensate demineralizer demineralizer (2) pipe tunnels 6Z26 15 mrem/hr III xxx 6Z27 III.5.6 Recycle evaporator 10Z35 Contaminated Pipe tunnel 6Z26 15 mrem/hr III xxx feed demineralizers demineralizer 6Z27 III.5.7 Cation bed 10Z32 Contaminated Pipe tunnels 6Z26 15 mrem/hr III xxx demineralizers demineralizer 6Z27 III.5.8 Mixed bed 10Z33 Contaminated Pipe tunnels 6Z26 15 mrem/hr III xxx demineralizers demineralizer 6Z27 12.3-66

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.5.9 Spent fuel pit 10Z34 Contaminated Pipe tunnels 6Z26 15 mrem/hr III xxx demineralizer demineralizer 6Z27 III.5.10 Radwaste evaporator 6Z5 Operation of Area outside 6Z1 2 mrem/hr I-D OA5-15A surface condensers 6Z6 evaporators surface condenser OA5-15B and feed pumps 6Z7 cubicles OA5-15C III.5.11 Boric acid tanks 6Z13 Contaminated Area outside tank 6Z1 2 mrem/hr I-D OA5-12 water cubicle OA5-13 III.5.12 Vertical pipe 6Z36 (1) pipes con- Area outside pipe 6Z2 2 mrem/hr I-D x(OA4-16B) taining waste tunnel gas sources 6Z33 (2) HVAC and Area outside 7Z1 2 mrem/hr I-D x(OA4-29) 6Z34 radioactive pipes pipe tunnel III.5.13 Laundry drain 6Z20 Normal operation Area outside tank 6Z1 2 mrem/hr I-D OA5-8 and laundry drain 6Z21 and filter tank filter III.5.14 Pipe tunnel 6Z26 Radioactive (1) area outside 6Z1 2 mrem/hr I-D 1A5-5 6Z27 water normal pipe tunnel 2A5-5 operation (2) valve aisle 6Z25 15 mrem/hr III xxx 6Z28 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after Area outside pipe 6Z1 100 mrem/hr - -

LOCA tunnel 12.3-67

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.5.15 Pipe penetration 6Z23 Normal operation Area outside 6Z1 2 mrem/hr I-D 1A5-1 areas 6Z32 radioactive penetration area 2A5-1 water in pipes 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after Area outside 6Z1 100 mrem/hr - -

LOCA III.5.16 Spent resin and 6Z3 (1) spent resins (1) outside pump 6Z1 2 mrem/hr I-D OP5-37 concentrates and evaporator cubicle 10 mrem/hr **

pumps concentrates (2) valve aisle 6Z4 15 mrem/hr III xxx (2) radioactive Valve operating 6Z1 2 mrem/hr I-D OA5-18 pipes and valve area III.5.17 Calibration 6Z30 Calibration (1) outside Room 6Z1 2 mrem/hr I-D OA5-20 rooms of instruments shielded OA5-21 OA5-22 (2) containment 6Z35 2 mrem/hr I-D x(OA5-20) roof stairs (3) interior 6Z38 4 mrem/hr II-A OA5-6 entrance hall III.6. Elevation 426 ft - 0 in.

III.6.1 Laundry room 5Z10 Operation of Area outside 5Z1 2 mrem/hr I-D OA6-9 laundry laundry room OA6-10 12.3-68

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP) 5Z10 Normal operation Area inside 5Z10 4 mrem/hr II-A OA6-8 (includes air- laundry room borne) 5Z11 Heavily contami- Hamper storage 5Z10 4 mrem/hr II-A OA6-7 nated laundry entrance III.6.2 Hot lab 5Z13 Processing Area outside 5Z1 2 mrem/hr I-D OA6-12 radioactive room OA6-19 samples III.6.3 Radwaste 5Z24 Operation of Area outside 5Z1 2 mrem/hr I-D OA6-24A evaporators 5Z26 evaporators evaporator OA6-24B 5Z27 cubicles OA6-24C III.6.4 Waste gas compressor 5Z21 Fission product Outside com- 5Z1 2 mrem/hr I-D OA6-25A packages 5Z22 gases pressor cubicle III.6.5 Automatic gas 5Z23 Analyzing (1) outside 5Z1 2 mrem/hr I-D OA6-25B analyzer fission product analyzer gases cubicle 5Z28 (2) outside 5Z1 2 mrem/hr I-D OA6-39 valve room III.6.6 Concentrates 5Z20 Storage of Area outside 5Z1 2 mrem/hr I-D OA6-26B holding tank evaporator cubicle concentrates III.6.7 Spent resin storage 5Z19 Radioactive (1) area outside 5Z1 2 mrem/hr I-D OA6-26A tank resins tank cubicle 12.3-69

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.6.8 Volume control tank 5Z6 Radioactive (1) area outside 5Z1 2 mrem/hr I-D 1A6-5 5Z7 water tank cubicle 2A6-5 (2) valve aisle 5Z8 15 mrem/hr III xxx 5Z9 5Z8 Pipe with Valve operating 5Z1 2 mrem/hr I-D 1A6-6 5Z9 radioactive water area 2A6-6 III.6.9 Mask cleaning room 5Z16 Decontamination Area outside 5Z1 2 mrem/hr I-D OA6-15A (Byron) and storage room III.6.9A Mask cleaning room 5Z32 Decontamination Area outside 5Z1 2 mrem/hr I-D OA6-13 (Braidwood) and storage room III.6.10 Decontamination 5Z31 Equipment decon- Area outside 5Z1 2 mrem/hr I-D OA6-4 Facility tamination and room storage III.7 Elevation 451 ft - 0 in.

III.7.1 Control room area 3Z7 (1) normal Inside control 3Z1 0.2 mrem/hr V OA7-2 3Z12 operation room area OA7-5 3Z14 TML (2) LOCA Inside control 3Z1 <5 rem during - -

direct plus room area the 30 days immersion dose post-LOCA 12.3-70

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

III.7.2 Purge room area 18Z10 Normal plant Inside purge room 3Z12 2 mrem/hr I-D 1A7-9 operation hot area between el. 3Z14 10 mrem/hr ** 2A7-9 spot applies to 451 ft. 0 in. and VQ penetrations 476 ft. - 6 in.

III.7.3 Auxiliary building 3Z13 Contaminated Corridor outside 3Z12 2 mrem/hr I-D OA7-8 HVAC charcoal filter VA charcoal charcoal filter 3Z14 area filters banks III.8 Areas separating main portion of auxiliary building from containment buildings III.8.1 Elevation 346 ft-0 in. TML (1) normal Outside separation 9Z1 2 mrem/hr I-D 1,2A2-12 Elevation 364 ft-0 in. and shutdown area in main 8Z1 1,2A3-10 Elevation 383 ft-0 in. operation portion of OA4-12 auxiliary building (2) radioactive Same as above 9Z1 100 mrem/hr - -

pipes 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 8Z1 after LOCA 7Z1 III.8.2 Elevation 401 ft-0 in. See III.5.16 III.8.3 Elevation 426 ft-0 in. See I.C.3 III.8.4 Elevation 439 ft-0 in. See I.C.3.

III.9 Auxiliary building TML Normal operation Auxiliary 1Z1 0.5 mrem/hr I-B 1,2A9-1 building roof 1,2A9-2 OA9-3 12.3-71

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

IV. Radwaste Building IV.1 Elevations 397'-0" and 401'-0" IV.1.1 Fluid bed dryer 19Z6 VR system Area outside 19Z2 2 mrem/hr I-D R5-10A processing waste room entrance 10Z22 10 mrem/hr ** R5-6 IV.1.2 Incinerator 19Z7 Incinerator Area outside 19Z12 2 mrem/hr I-D R5-12 processing dry room 10 mrem/hr **

active waste IV.1.3 Scrubber and 19Z4 VR system Area outside 19Z12 2 mrem/hr I-D R5-13 feed pumps 19Z13 processing waste cubicles IV.1.4 Feed tank 19Z10 Radioactive Area outside 19Z2 2 mrem/hr I-D R5-10A recirculation pumps sludge and water pump cubicle IV.1.5 Radwaste drumming 19Z9 Waste filled Area outside 19Z22 2 mrem/hr I-D R5-6 station drum drumming cubicle IV.1.6 Drum swipe and 19Z23 Waste filled Area in front 19Z14 2 mrem/hr I-D R5-4 labeling station drum of station 10 mrem/hr **

IV.1.7 Drum storage 19Z27 Waste filled (1) truck bay 19Z22 2 mrem/hr I-D R5-7 areas 19Z26 drums R5-9 (2) loading 19Z14 2 mrem/hr I-D R5-5 platform 12.3-72 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

IV.1.8 Truck bay 19Z22 Loading Waste (1) Radwaste 19Z14 2 mrem/hr I-D R5-1 filled drums bldg entrance (2) RB control 19Z16 1 mrem/hr I-C R5-2 room IV.2 Elevation 410 ft - 0 in.

IV.2.1 Fluid bed dryer 19Z6 VR system Area outside 20Z1 0.2 mrem/hr I-A S6-3 processing waste room IV.2.2 Drumming station 19Z9 Waste filled Area above 20Z1 0.2 mrem/hr I-A S6-5 drum drumming unit IV.2.3 Feed tanks 19Z1 Radioactive Pump room 19Z10 15 mrem/hr III ***

sludge and water entrance IV.2.4 Gas/Solid separator 19Z5 VR system in Operator area 19Z3 4 mrem/hr II-A R6-3 operation IV.2.5 VR system charcoal 19Z8 VR system in Entrance to 19Z3 4 mrem/hr II-A R6-4 filter operation incinerator room IV.2.6 Recirculation 19Z10 Normal Shredder 19Z2 2 mrem/hr I-D R5-10B pump skid operation area IV.2.7 Transfer product 19Z19 Normal Area outside 19Z12 2 mrem/hr I-D R5-14 hopper cubicle operation cubicle IV.2.8 Dryer feed TML Normal Area outside 19Z12 2 mrem/hr I-D R5-15 tunnel operation tunnel 12.3-73

B/B-UFSAR TABLE 12.3-2 (Cont'd)

RADIATION REFERENCE SOURCE PROTECTED DESIGN DOSE ZONE BASE POINT NUMBER LOCATION OR SOURCE ZONE NO. DESIGN CONDITION PROTECTED AREA ZONE NO. RATE* DESIGNATION (RBP)

V. Turbine building V.1 Turbine building TML Normal operation Inside building TML 1 mrem/hr I-C 1,2T5-3 V.2 Safety valve en- 16Z5 Normal operation Inside enclosure 16Z7 4 mrem/hr II-A -

closure to 16Z6 16Z8 containment V.3 Condensate polishing 21Z1 When polishers Turbine bldg. TML 1 mrem/hr I-C OT5-7 area are used doorway Notes to Table

1. The zone designations are discussed in Table 12.3-1.

2. The "Zone Numbers" are shown on Figures 12.3-5 through 12.3-26 Drawings M-24-1 through M-24-23.

3. The design dose rate values given in this table are based on the design criteria for a solid shielding unit and does not reflect the impact of penetrations and voids except for a few protected areas that have ** indicated in the "Zone Designation" column. The exceptions indicate a few select hot spots, but this criteria can be applied to every protected area listed above. A detailed explanation of the hot spot criteria can be found in Subsection 12.3.2.3.

4. Verification of shielding walls and slabs that separate two radiation areas is not practical because the radiation fields(s) coming through the wall during normal operation will be masked by the radiation field in the protected area. Therefore, it is not practical to have a radiation base point for the area. The design dose rate is specified to protect maintenance operations. The "xx" and "xxx" in the radiation base point column identify these special types of protected areas.

12.3-74 REVISION 9 - DECEMBER 2002

B/B-UFSAR TABLE 12.3-3 AREA RADIATION MONITORS RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS ORE-AR001 Aux. Bldg. El. 346 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR002 Aux. Bldg. El. 346 0.1-10,000 mR/hr GM 0.08-3 MeV per RP approved procedures ORE-AR003 Aux. Bldg. El. 346 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR004 Aux. Bldg. El. 364 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR005 Aux. Bldg. El. 364 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR006 Aux. Bldg. El. 364 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR007 Aux. Bldg. El. 383 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR008 Aux. Bldg. El. 383 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR009 Aux. Bldg. El. 383 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR010 Aux. Bldg. El. 401 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR011 Aux. Bldg. El. 401 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR012 Aux. Bldg. El. 401 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR013 Aux. Bldg. El. 401 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR014 Aux. Bldg. El. 426 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR015 Aux. Bldg. El. 426 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

12.3-75 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS ORE-AR016 Aux. Bldg. El. 426 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR017 Aux. Bldg. El. 451 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures ORE-AR031 Primary Sample Room 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

ORE-AR032 High Level Lab El. 426 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

ORE-AR035 Drumming Station 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 383 ORE-AR037 Fuel Handling Bldg. 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 426 ORE-AR038 Fuel Handling Bldg. 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 ORE-AR039 Fuel Handling Bldg. 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures* Interlock Crane Trolley Crane Raise El. 426 Circuit ORE-AR041 Radwaste Bldg. Low 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

Level Storage El. 410 ORE-AR042 Radwaste Bldg. El. 401 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

ORE-AR043 Radwaste Bldg. Truck 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

Bay El. 397 ORE-AR044 Radwaste Bldg. Low 1-100,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

Level Storage El. 401 12.3-76 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS ORE-AR045 Radwaste Bldg. High 1-100,000 mR/hr GM 0.08-3 MeV per RP-approved procedures* High back-Level Storage El. 401 ground area ORE-AR046 Volume Reduction Area 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 ORE-AR047 Volume Reduction Area 1-100,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 ORE-AR048 Volume Reduction Area 1-100,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 ORE-AR049 Volume Reduction Area 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 ORE-AR050 Volume Reduction Area 1-100,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 1RE-AR001 Containment El. 426 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

2RE-AR001 Containment El. 426 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

1RE-AR002 Containment El. 401 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

2RE-AR002 Containment El. 401 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

1RE-AR003 Incore Seal Table 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 2RE-AR003 Incore Seal Table 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 401 12.3-77 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS 1RE-AR010 Main Control Room 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures El. 451 2RE-AR010 Main Control Room 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures El. 451 1RE-AR011 Containment Fuel 0.1-10,000 mR/hr GM 0.08-3 MeV 2 x background Redundant Handling Incident in the Containment with El. 426 Building at RTP 1RE-AR012 2RE-AR011 Containment Fuel 0.1-10,000 mR/hr GM 0.08-3 MeV 2 x background Redundant Handling Incident in the Containment with El. 426 Building at RTP 2RE-AR012 1RE-AR012 Containment Fuel 0.1-10,000 mR/hr GM 0.08-3 MeV 2 x background Handling Incident in the Containment El. 426 Building at RTP 2RE-AR012 Containment Fuel 0.1-10,000 mR/hr GM 0.08-3 MeV 2 x background Handling Incident in the Containment El. 426 Building at RTP 0RE-AR055 Fuel Building Fuel 0.1-10,000 mR/hr GM 0.08-3 MeV 5 mR/hr Redundant Handling Incident with El. 426 0RE-AR056 0RE-AR056 Fuel Building Fuel 0.1-10,000 mR/hr GM 0.08-3 MeV 5 mR/hr Handling Incident El. 426 12.3-78 REVISION 15 - DECEMBER 2014

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS 1RE-AR013 Volume Control Tank 0.1-10,000 mR/hr IC 0.08-3 MeV per RP-approved procedures* High back-Cubicle El. 426 ground cubicle 2RE-AR013 Volume Control Tank 0.1-10,000 mR/hr IC 0.08-3 MeV per RP-approved procedures* High back-Cubicle El. 426 ground cubicle 1RE-AR020 High Range Containment 100-108 R/hr IC Per E-Plan EALs El. 514'-8" (Actual detector El.)

2RE-AR020 High Range Containment 100-108 R/hr IC Per E-Plan EALs El. 514'-8" (Actual detector El.)

1RE-AR021 High Range Containment 100-108 R/hr IC Per E-Plan EALs El. 514'-8" (Actual detector El.)

2RE-AR021 High Range Containment 100-108 R/hr IC Per E-Plan EALs El. 514'-8" (Actual detector El.)

ORE-AR073 TSC Monitor Room 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

El. 435 ORE-AR074 TSC Health Physics 0.1-10,000 mR/hr GM 0.08-3 MeV per RP-approved procedures*

Office El. 451 1RE-AR022A Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 1A with 1RE-AR023A 12.3-79 REVISION 14 - DECEMBER 2012

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS 1RE-AR022B Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 1B with 1RE-AR023B 1RE-AR022C Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 1C with 1RE-AR023C 1RE-AR022D Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 1D with 1RE-AR023D 2RE-AR022A Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 2A with 2RE-AR023A 2RE-AR022B Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 2B with 2RE-AR023B 2RE-AR022C Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 2C with 2RE-AR023C 2RE-AR022D Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background Redundant 2D with 2RE-AR023D 1RE-AR023A Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 1A 1RE-AR023B Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 1B 12.3-80 REVISION 14 - DECEMBER 2012

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS 1RE-AR023C Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 1C 1RE-AR023D Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 1D 2RE-AR023A Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 2A 2RE-AR023B Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 2B 2RE-AR023C Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 2C 2RE-AR023D Main Steamline 0.1-10,000 mR/hr GM 0.02-3 MeV 3X background 2D 1RE-AR024A Main Steamline 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures 1A & 1D Pen. R13 1RE-AR024B Main Steamline 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures 1B & 1C Pen. R20 2RE-AR024A Main Steamline 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures 2A & 2D Pen. R41 2RE-AR024B Main Steamline 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures 2B & 2C Pen. R34 1RE-AR025A Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 364' - R5 12.3-81 REVISION 14 - DECEMBER 2012

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS 1RE-AR025B Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 364' - R7 2RE-AR025A Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 364' - R28 2RE-AR025B Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 364' - R26 1RE-AR026A Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 383' - R5 1RE-AR026B Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 383' - R7 2RE-AR026A Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 383' - R28 2RE-AR026B Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 383' - R26 1RE-AR027A Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 401' - R5 1RE-AR027B Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 401' - R7 12.3-82 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.3-3 (Cont'd)

RADIATION TYPE OF DETECTOR NO SERVICE RANGE DETECTOR ENERGY RANGE SETPOINT REMARKS 2RE-AR027A Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 401' - R28 2RE-AR027B Piping Penetration 0.1-10,000 R/hr IC 0.1-3 MeV per RP-approved procedures El. 401' - R26

Local indication and alarm provided.

12.3-83 REVISION 7 - DECEMBER 1998

B/B-UFSAR TABLE 12.3-4 PARAMETERS USED IN THE CALCULATION OF THE PRIMARY SHIELD THICKNESS CORE POWER RATING Total Core Thermal (Mw) 3565 Power Density (watts/cc) 109.2 CORE EFFECTIVE DIMENSIONS (cm)

Height 365.76 Diameter 337.09 CORE VOLUME FRACTIONS UO2 .3052 Zirconium .0943 Stainless Steel .0053 Inconel .0043 Water .5909 REACTOR DIMENSIONS OUTSIDE REGION MATERIAL RADIUS (cm) THICKNESS (cm)

Core (see above) 168.545 168.545 Baffle SS 172.402 2.858 Water H2O 187.960 16.558 Barrel SS 193.675 5.715 Shield Panel Void* 200.667 6.985 Water H2O 219.71 19.05 Pressure Vessel CS 241.618 101.7 Void + Neutron - - -

Detector Cavity Air 343.318 -

Primary Shield Ordinary concrete 517.208 173.89

The worst case radial traverse was chosen for the ANISN model which travels in-between the intermittent shield panels and goes through a neutron detector cavity.

12.3-84 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.3-4 (Cont'd)

CORE RADIAL SOURCE DISTRIBUTION FRACTION OF RADIAL POWER CORE RADIUS DISTRIBUTION

.1 1.0

.2 0.999

.3 0.998

.4 0.996

.5 0.992

.6 0.975

.7 0.942

.8 0.858

.9 0.67 1.0 0.563 12.3-85

B/B-UFSAR TABLE 12.3-5 CORE FISSION SOURCE FOR PRIMARY SHIELD CALCULATION UPPER CORE TOTAL GROUP ENERGY (MeV) NEUTRON SOURCE (n/cc-sec) 1 15 1.32 x 109 2 12.2 7.56 x 109 3 10.0 2.94 x 1010 4 8.18 1.25 x 1011 5 6.36 2.86 x 1011 6 4.96 4.09 x 1011 7 4.06 9.08 x 1011 8 3.01 7.53 x 1011 9 2.46 1.98 x 1011 10 2.35 1.02 x 1012 11 1.83 1.85 x 1012 12 1.11 1.68 x 1012 13 0.55 1.15 x 1012 14 0.111 1.31 x 1011 15 0.003 0.0 TOTAL 8.53 x 1012 12.3-86

B/B-UFSAR TABLE 12.3-6 SHIELDING DESIGN-BASIS GEOMETRY FOR SHIELDING THICKNESS CALCULATIONS HOMOGENIZED COMPUTER SOURCE SOURCE NAME OF CODE SOURCE DENSITY DIMENSIONS***

COMPONENT SOURCE USED* COMPOSITION (gm/cc) GEOMETRY** (ft)

Reactor fission A UO2, Zr, SS 4.4 C R = 5.5 (core) spectrum Inconel, H2O Steam Generator Table 12.2-1, I H2O, Fe .756 C R = 11.3 12.2-2, 12.2-3 H = 28.3 Pressurizer Table 12.2-4, I H2O .68 C R = 3.5 12.2-5, 12.2-6 H = 38 (normal water level)

Reactor Coolant Pumps Table 12.2-1, I H2O .68 C R = 1.25 and Piping 12.2-2 H = as required Reactor Coolant Drain Table 12.2-2 I H2O 1.0 C R = 1.5 Tank H = 7.4 Regenerative Heat Table 12.2-1 I H2O 1.0 C R = .83 Exchanger H = 18 Excess Letdown Heat Table 12.2-1 I H2O 1.0 C R = .75 Exchanger H = 14 Incore Detectors and Table 12.2-26, I Fe 7.87 L H = 15 Drive Wires 12.2-27 12.3-87 REVISION 3 - DECEMBER 1991

B/B-UFSAR TABLE 12.3-6 (Cont'd)

HOMOGENIZED COMPUTER SOURCE SOURCE NAME OF CODE SOURCE DENSITY DIMENSIONS***

COMPONENT SOURCE USED* COMPOSITION (gm/cc) GEOMETRY** (ft)

Fuel Assembly in Table 12.2-23 I UO2, Zr, SS 4.4 S W = .7 Refueling Cavity adjusted for one Inconel, H2O L = .7 fuel assembly H = 13.25 (4-day decay)

Volume Control Tank Table 12.2-8 I H2O liquid: 1.0 R = 3.25, H = 3.6 vapor: 0.001293 C R = 3.25, H = 5.4 Recycle Holdup Tank Table 12.2-9, I H2O liquid: 1.0 R = 14, H = 12.5 12.2-10 vapor: 0.001293 R = 14, H = 14.5 Recycle Evaporator Table 12.2-11 I H2O 1.0 C R = 1.8 H = 9.9 Recycle Evaporator Table 12.2-11 I H2O 1.293E-3 C R = .33 Vent Condenser H = .8 RHR Heat Exchanger Table 12.2-12 I H2O 1.0 C R = 3.6 H = 28 RHR Pump and Piping Table 12.2-12 I H2O 1.0 C R = .58 H = 17 Mixed Bed Table 12.2-13 I/M* H2O 1.0 C R = 1.083 Demineralizer H = 8 Cation Bed Table 12.2-14 I H2O 1.0 C R = 1.33 Demineralizer H = 3.6 Thermal Regeneration Table 12.2-15 I H2O 1.0 C R = 1.0 Demineralizer H = 5.6

For Braidwood 12.3-88 REVISION 14 - DECEMBER 2012

B/B-UFSAR TABLE 12.3-6 (Cont'd)

HOMOGENIZED COMPUTER SOURCE SOURCE NAME OF CODE SOURCE DENSITY DIMENSIONS***

COMPONENT SOURCE USED* COMPOSITION (gm/cc) GEOMETRY** (ft)

Recycle Evaporator Table 12.2-16 I H2O 1.0 C R = 1.083 Feed Demineralizer H = 8 Recycle Evaporator Table 12.2-17 I H2O 1.0 C R = 1.083 Condensate H = 8 Demineralizer Spent Fuel Pit Table 12.2-18 I H2O 1.0 C R = 1.083 Demineralizer H = 8 Reactor Coolant Filter Table 12.2-19 I H2O .38 C R = .28 H = 1.6 Seal Water Return Table 12.2-20 I H2O .38 C R = .28 Filter H = 1.6 Recycle Evaporator Table 12.2-20 I H2O .38 C R = .28 Feed Filter H = 1.6 Spent Fuel Pit Filter Table 12.2-20 I H2O .38 C R = .28 H = 1.6 Spent Fuel Pit Skimmer Table 12.2-20 I H2O .38 C R = .28 Filter H = 1.6 Seal Water Injection Table 12.2-20 I H2O .38 C R = .11 Filter H = 1.7 Recycle Evaporator Table 12.2-21 I H2O .38 C R = .11 Concentrates Filter H = 1.7 12.3-89

B/B-UFSAR TABLE 12.3-6 (Cont'd)

HOMOGENIZED COMPUTER SOURCE SOURCE NAME OF CODE SOURCE DENSITY DIMENSIONS***

COMPONENT SOURCE USED* COMPOSITION (gm/cc) GEOMETRY** (ft)

Recycle Evaporator Table 12.2-22 I H2O .38 C R = .1 Condensate Filter H = 1.7 Waste Gas Decay Tanks Table 12.2-22 I H2O 0.001293 cylindrical R = 4.25 H = 10.6 Spent Fuel Storage 5/3 core fission I UO2, Zr, SS 2.75 S W = 11.88 Area products Inconel, H2O L = 16.25 H = 62 (Transfer of) One Table 12.2-23 Q UO2, Zr, SS 4.4 S W = .7 Spent Fuel Assembly adjusted for one Inconel, H2O L = .7 fuel assembly H = 13.25 (4-day decay)

Laundry Drain Tank Table 12.2-33 I H2O 1.0 C R = 3.25 H = 16.5 Blowdown Mixed Bed Table 12.2-35 I H2O 1.0 C R = 2 Demineralizer H = 5.16 Radwaste Mixed Bed Table 12.2-35 I H2O 1.0 C R = 2 Demineralizer H = 5.16 Concentrates Holding Table 12.2-36 I H2O 1.0 C R = 5 Tank H = 8.5 Blowdown Prefilter Table 12.2-37 I H2O 1.0 C R =

H =

12.3-90 REVISION 5 - DECEMBER 1994

B/B-UFSAR TABLE 12.3-6 (Cont'd)

HOMOGENIZED COMPUTER SOURCE SOURCE NAME OF CODE SOURCE DENSITY DIMENSIONS***

COMPONENT SOURCE USED* COMPOSITION (gm/cc) GEOMETRY** (ft)

Blowdown Afterfilter Table 12.2-37 I H2O 1.0 C R = .25 H =

Radwaste Afterfilter Table 12.2-37 I H2O 1.0 C R = .25 H =

Turbine Building Table 12.2-37 I H2O 1.0 C R = .25 Equipment Drain Filter H =

Turbine Building Table 12.2-37 I H2O 1.0 C R = .25 Floor Drain Filter H = 1.6 Auxiliary Building Table 12.2-38 I H2O 1.0 C R = .25 Equipment Drain Filter H = 1.6 Auxiliary Building Table 12.2-38 I H2O 1.0 C R = .25 Floor Drain Filter H = 1.6 Regeneration Waste Table 12.2-38 I H2O 1.0 C R = .25 Drain Filter H = 1.6 Chemical Drain Filter Table 12.2-38 I H2O 1.0 C R = .25 H = 1.6 Laundry Drain Filter Table 12.2-38 I H2O 1.0 C R = .25 H = 1.6 Radwaste Evaporator Table 12.2-39 I H2O 1.0 C R = 3 H = 15 12.3-91

B/B-UFSAR TABLE 12.3-6 (Cont'd)

HOMOGENIZED COMPUTER SOURCE SOURCE NAME OF CODE SOURCE DENSITY DIMENSIONS***

COMPONENT SOURCE USED* COMPOSITION (gm/cc) GEOMETRY** (ft)

Radwaste Evaporator I H2O 1.0 C R = 1.5 Surface Condenser H = 11 30,000 gal. Release Table 12.2-40 I H2O 1.0 C R = 17 Tank H = 8.6 Permeate Sample Tank Table 12.2-40 Blowdown Monitor Tank Table 12.2-41 I H2O 1.0 C R = 8 H = 15.83 Radwaste Evaporator Table 12.2-41 I H2O 1.0 C R = 8 Monitor Tank H = 15.83 Spent Resin Tank Table 12.2-43, I H2O 1.2 C R = 4.5 col. 1 H = 10.5 Radwaste Drum Storage Table 12.2-43, I Table 12.2-44 1.33 S W = 14.66 col. 2 L = 18.33 H = 15 Refueling Water Table 12.2-25 I H2O 1.0 C R = 25 Storage Tanks Concrete 2.242 H = 30 Condensate Storage 10-3 Ci/cc I H2O 1.0 C R = 22 Tank @1.3 MeV H = 44 12.3-92

B/B-UFSAR TABLE 12.3-6 (Cont'd)

HOMOGENIZED COMPUTER SOURCE SOURCE NAME OF CODE SOURCE DENSITY DIMENSIONS***

COMPONENT SOURCE USED* COMPOSITION (gm/cc) GEOMETRY** (ft)

Auxiliary Building Table 12.2-45 I C .45 S W = 2.4 Charcoal Filters L = 16.75 H = 7.5 Steam Jet Air Ejector Table 12.2-45 I C .45 S W = 2.4 Vent Filter System L = 16.75 H = 7.5

A = ANISN, I = ISOSHLD, Q = QAD, M=Microshield

- C = cylindrical, S = finite slab, L = line

- - R = radius, H = height, L = length, W = width

- - - The permeate sample tank sources are less than or equal to the laundry drain tank sources.

Thus, the same shielding requirement was recommended.

Shielding was determined based on equipment 1/2WX02MA,B (housing-only prefilter vessels).

12.3-93 REVISION 14 - DECEMBER 2012

B/B-UFSAR TABLE 12.3-7 ESTIMATED OCCUPATIONAL RADIATION EXPOSURE DURING DECOMMISSIONING ALTERNATIVE MAN-REM Mothballing 150 Entombment 130 Prompt Dismantling 630 Mothballing with Delayed Dismantling* 150 + 310 Entombment with Delayed Dismantling* 130 + 310

104-year delay period before delayed dismantling.

The above information was assembled from Reference 10.

12.3-94

B/B-UFSAR TABLE 12.3-8 DOMINANT RADIOACTIVE ISOTOPES FOR PROMPT DISMANTLING AND DELAYED DISMANTLING PROMPT DISMANTLING DELAYED DISMANTLING SOURCE 2 YEARS OF DECAY 104 YEARS OF DECAY Vessel and Internals Fe55, Co60, Ni63 Ni63 Other systems Co60 Sr90, Cs137 The above information was assembled from Reference 10.

12.3-95

B/B-UFSAR TABLE 12.3-9 SENSITIVITY OF CONTINUOUS AIRBORNE MONITORING SYSTEM AVERAGE GROSS SENSITIVITY ENERGY SENSITIVITY (cpm/hr MPC*** DETECTABILITY*

ISOTOPE (MEV) (cpm/Ci) per Ci/cc) (Ci/cc) FACTOR FOR MPCa I. Particulate Channel (Beta Scintillator)

CO-60 0.096 4.69 x 105 2.01 x 1012 9 x 10-9 200 6

SR-90 0.200 1.20 x 10 5.10 x 1012 1 x 10-9 60 TC-99 0.085 4.47 x 105 1.92 x 1012 6 x 10-8 1350 CS-137 0.171 1.15 x 106 4.91 x 1012 1 x 10-8 575 II. Iodine Channel (NaI Spectrometry windowed on I-131 peak)

I-131 0.364() 1.01 x 105 4.29 x 1011 9 x 10-9 850 III. Noble Gases Channel (Beta Scintillator)

KR-85 0.100 ----- 1.84 x 107** 1.0 x 10-5 40 XE-133 0.250 ----- 3.6 x 107** 1.0 x 10-5 80

The Minimum detectable (activity) concentration is based on a signal count rate at a 95% confidence level as given by the formula in ANSI 13.10-1974 and modified for the GA system as follows:

MDC = 2 (BCKG/20)1/2 Sensitivity, (for BCKG 100 cpm) 2 (BCKG2/2000)1/2 Sensitivity, (for 100 cpm < BCKG < 1 x 105 cpm)

Where BCKG is the total background counting rate (cpm). For the particulates 1905 cpm was used and for iodine and noble gases 100 cpm was used; this criterion will yield an answer that has a 95% statistical confidence level.

- cpm per Ci/cc

- - The term MPC refers to a 10CFR20 limit in effect prior to January 1, 1994.

12.3-96 REVISION 8 - DECEMBER 2000

B/B-UFSAR ATTACHMENT 12.3A EXAMPLES OF THE APPLICATION OF RADIATION PROTECTION DESIGN FEATURES TO SPECIFIC COMPONENTS 12.3A-1

B/B-UFSAR EXAMPLES OF THE APPLICATION OF RADIATION PROTECTION DESIGN FEATURES TO SPECIFIC COMPONENTS The general principles and concepts of radiation protection design features including shielding to minimize occupational dose are described in the various subsections of 12.3.

The application of these features to the design of specific components is described below.

DEMINERALIZERS The demineralizers are isolated from their valves, other equipment, and from general access areas. In addition to labyrinth entrances, some demineralizer rooms have removable ceiling hatches. At least one hatch contains a radiation probe hole which is utilized prior to removing the hatch.

The metering device attached to the probe is properly calibrated so that operating personnel will have adequate radiation data prior to removing the hatch.

The valves for the demineralizers are located in a separate room. A typical arrangement is shown in Figure 12.3A-1. Valve operator stations located in general access areas are utilized wherever they are practical. Ventilation to the valve room is supplied from the general access area and is exhausted to the demineralizer room and/or the radwaste tunnel. The ventilation exhaust from the demineralizer room goes directly into an adjacent radwaste pipe tunnel.

SAMPLING STATION Sampling stations can be located singly (inside labyrinth entrances when practical) or can be grouped together in sample panels. The sampling station is located as close to the sampling point as is practical, but not in direct view of a radioactive source. Shielding, drains, and flushing lines are used to reduce occupational radiation exposure whenever it is practicable to do so. A single sampling station and a sample panel are shown in Figure 12.3A-1.

HYDROGEN RECOMBINER The hydrogen recombiner is a postaccident system. The containment hydrogen recombiners are located in a general access area at elevation 401 feet 0 inch adjacent to column rows 15U and 21U. This location was selected so that each recombiner is close to the containment yet shielded from it. The radiation shielding surrounding the recombiner is designed to protect the area directly adjacent to the recombiner from the postaccident radiation sources and to allow access to the recombiner (for maintenance, removability, and replacement) during the postaccident period as well as during normal station operation.

12.3A-2

B/B-UFSAR The recombiners are only to be operated during postaccident conditions and when they are being tested. Therefore, the recombiners will not become radioactive during normal station operation. A removal fence may be used to keep the recombiner removal path clear of traffic and equipment.

Start switches for the recombiners are located in the recombiner controls console. The Unit 1 recombiner control console is located away from the recombiners on elevation 401 feet 0 inch (column row 13/P). The Unit 2 recombiner control console is located away from the recombiners on elevation 439 feet 0 inch (column row 25/Q).

Area radiation monitors (ARMs) are located near the recombiner area so that station personnel will be alerted to high radiation levels. The Unit 1 ARM and recombiner area are shown on Figure 12.3A-2.

EVAPORATORS The radioactive evaporator equipment is segregated from the remainder of the evaporator equipment. The radioactive equipment is located on an upper level which has only one access (a shielded staircase). Access to the upper level is through a closed door which is utilized in accordance with 10 CFR 20. The lower level contains the evaporator condensing equipment, the radiation monitor panel, and the control panel. This equipment is slightly radioactive (approximately 1 x 10-4 times the dose rate of the upper level) and needs to be separated from general access areas. Figure 12.3A-3 shows the layout of the evaporator equipment.

FUEL TRANSFER TUBE The fuel transfer area is shown in Figures 12.3A-4, and 12.3A-5, and Drawing M-24 Sheet 14 and 16.

The shielding for the fuel transfer tube is based on a peak fuel assembly. This is an assembly that has 1.5 times the average 1000-day burnup. In order to obtain a dose rate of 5 mrads/hr in adjacent areas, 5 feet of ordinary concrete is required. The radiation streaming through the 2-inch expansion gap is reduced by attaching a 5-inch thick, 3-inch wide steel horseshoe shielding collar on the transfer tube sleeve.

The expected doses during fuel transfer are:

elevation 389 < 5 mrem/assembly, elevation 399 < 5 mrem/assembly, and tendon tunnel < 2 mrem/assembly.

12.3A-3 REVISION 9 - DECEMBER 2002

B/B-UFSAR There will be zero access to the fuel transfer tube area during periods when fuel is being moved through the tube.

The fuel transfer tube will only be exposed for tube inspection.

These inspections will only be scheduled for times when no fuel movement is scheduled. The tube inspection and the replacing of all shielding that was removed will be performed on a priority basis. The entire operation will be completed in the shortest time practical. Key operating personnel, especially the fuel transfer office are informed at the beginning and at the completion of the tube inspection.

12.3A-4

B/B-UFSAR 12.4 DOSE ASSESSMENT 12.4.1 Estimated Occupancy of Plant Radiation Zones It is difficult to predict the occupancy of any one zone on an average weekly basis, much less by function. An estimate based on an average yearly occupancy (even if such data existed) would give questionable results in dose calculations because the exposure is related to the operating history, which is highly episodic.

12.4.2 Estimates of Inhalation Doses Small airborne radioactivity concentrations of radionuclides are expected within the various plant structures. Implementation of the health physics program (Section 12.5) mitigates against any significant ingestion doses to plant personnel.

Radionuclides are a potential hazard because they may be present and released in significant quantities from fluids. For this reason, maximum design-basis radioiodine concentrations have been calculated in the buildings most susceptible to airborne contamination. The assumptions used in calculating these air-borne radioactivity concentrations are presented in Subsection 12.2.2.3. The resultant design-basis calculated concentrations are also tabulated in Tables 12.2-46, 12.2-47, and 12.2-48, and the thyroid dose acquired from their inhalation is tabulated in Table 12.4-4.

12.4.3 Objectives and Criteria for Design Dose Rates The objectives and criteria for design dose rates in various areas are discussed in Section 12.3.

12.4.4 Estimated Annual Occupational Exposures Many ALARA features have been incorporated into the design of Byron/Braidwood Stations - selection of materials to reduce crud levels, separation of radiation areas, etc. The actual dose rates experienced in most areas are expected to be lower than the respective design dose rates. So a calculation of occupational exposures based on design dose rates would overestimate these exposures in most cases. But either actual or design dose rate calculations would require a knowledge of occupancy factors.

Thus the most reliable prediction of occupational exposures must be based on data from the operating history of other PWRs. Table 12.4-1 shows the annual personnel exposure reported for various PWRs. Much of the reported data for older plants is due to backfitting work, while the multiple-unit plants are not strictly comparable to single-unit plants, due to shared facilities such as radwaste areas. Table 12.4-2 shows reported 12.4-1 REVISION 8 - DECEMBER 2000

B/B-UFSAR occupational exposure for six younger multiple-unit PWRs.

Averages of total annual occupational exposure and average exposure per person differ in the two tables, as expected.

Table 12.4-2 also shows estimates of annual occupational exposure based on the average values of the total annual occupational exposure and average exposure per person. For older single-unit PWRs, the annual occupational exposure varies from year to year, but generally lies in a range between 60 and 1000 person-rem after the first year (Reference 5). Since Byron/Braidwood Stations are multiple-unit plants, it would be reasonable to assume that the annual occupational exposure is in the upper part of that range. Considering the experience at Zion during the first 4 years of operation, however, 800 to 1200 person-rem is probably an upper bound on the exposure. Thus, an estimate of 800 person-rem is given as the average annual occupational exposure.

Table 12.4-3 gives the reported occupational personnel exposure by work function for various operating plants. For the sake of comparison, data for three mature single-unit PWRs are shown.

The estimated annual occupational exposures by work function, also given in Table 12.4-3, are based on the averages for the multiple-unit plants.

Table 12.4-5 compares the occupational exposure for single unit stations, two-unit stations, and Zion. The fifth year radiation dose estimate for Byron/Braidwood Stations is assumed to have the same exposure as the average two-unit station for routine operations and for special maintenance. Byron/Braidwood Stations' exposure will improve on Zion in the areas of routine maintenance and radwaste (see Section 12.3). Byron/Braidwood Stations' refueling exposure is assumed to be slightly above that of Zion because of greater fuel burnup.

12.4.5 Estimated Annual Dose at the Exclusion Area Boundary The estimated annual dose at the exclusion area boundary (EAB) is given in Subsection 5.2.4 of the environmental report.

12.4.6 Deleted 12.4.7 Dose Reduction Program Commonwealth Edison utilizes an alternate empirical feedback method of design refinements to reduce radiological exposure and maintain doses ALARA. The areas of greatest total exposure in person-rem are identified, then design feedback is made to ameliorate the conditions such that actual exposure rate in the newer plant will be decreased in similar areas.

12.4-2 REVISION 8 - DECEMBER 2000

B/B-UFSAR An independent review was performed by Commonwealth Edison's corporate radiation protection group, which supports the generating stations. These production support specialists accumulated radiation histories from all of Commonwealth Edison's nuclear stations, and they also evaluated operations and procedures. Based on this data, the AE made the following design changes:

a. a partially mechanized solid radwaste system,

b. probe holes for radiation monitoring of filter cubicles,

c. lifting aids, and

d. improved drain systems.

Additional design features which were made to implement ALARA are described in Subsection 12.1.2.3.

The above process of dose improvement continues into the operation phase where the radiation evaluation program identifies areas where improvements are needed. Engineering support to the station is continuous, and these same feedback mechanisms continue to emphasize ALARA doses.

12.4.8 Radiological Environmental Monitoring Program The radiological environmental monitoring program (REMP) conducted in the vicinity of the stations has as its objectives:

a. Provide data on measurable levels of radiation and radioactivity in the environment and relate these data to radioactive emissions;

b. Identify changes in the use of nearby offsite areas to ensure adequate surveillance and evaluation of doses to individuals from principal pathways of exposure;

c. Provide environmental surveillance in case of an unplanned release; and

d. Provide year-round monitoring of principal pathway exposure.

The REMP provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of members of the public resulting from the station operation. This monitoring program implementsSection IV.b.2 of Appendix I to 10 CFR 50 and, thereby, supplements the 12.4-3 REVISION 8 - DECEMBER 2000

B/B-UFSAR radiological effluents monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of effluent measurements and the modeling of the environmental exposure pathways.

The site specific annex of the Offsite Dose Calculation Manual describes the current REMP and presents the required detection capabilities for environmental sample analyses tabulated in terms of the a priori minimum detectable concentration (MDC). The a priori MDC is a before-the-fact limit representing the capabilities of a measurement system and is not an after the fact limit for a particular measurement.

12.4.9 References

1. NUREG-0109, "Occupational Radiation Exposure at Light Water Cooled Power Reactors, 1969-1975," USNRC, August 1976.

2. "Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants," Atomic Industrial Forum, September 1974.

3. Northern States Power Company, Prairie Island Nuclear Generating Plant, Units 1 and 2, Semiannual Operating Report Number 4, January to June 1975, Docket No. 50282-401, August 1975.

4. Northern States Power Company, Prairie Island Nuclear Generating Plant, Units 1 and 2, Semi-Annual Operating Report Number 5, July to December 1975, Docket No. 50282-493, March 1976.

5. Rochester Gas and Electric Corporation, Ginna Nuclear Power Plant Annual Operating Report Number 11, January 1, 1975 through December 31, 1975, Docket No. 50244-497, February 1976.

6. Connecticut Yankee Atomic Power Company, Haddam Neck Plant Annual Operating Report, 1975, Docket No. 50213-484.

12.4-3a REVISION 5 - DECEMBER 1994

B/B-UFSAR

7. Maine Yankee Atomic Power Company, Maine Yankee Atomic Power Plant Semiannual Occupational Radiation Exposure Report, Errata, January to June 1975, Docket No. 50309-858, March 1976.

8. Maine Yankee Atomic Power Company, Maine Yankee Atomic Power Plant Semiannual Operating Report, July to December 1975, Docket No. 50309, March 1976.

9. Virginia Electric and Power Company, Surry Power Station, Units 1 and 2, Semiannual Operating Report, July to December 1974, Docket No. 50280, August 1975.

10. Virginia Electric and Power Company, Surry Power Station, Units 1 and 2, Semiannual Operating Report, January to June 1975, Docket No. 50280, August 1975.

11. Virginia Electric and Power Company, Surry Power Station, Units 1 and 2, Semiannual Operating Report, July to December 1975, Docket No. 50280, March 1976.

12. Commonwealth Edison Company, Zion Station, Units 1 and 2, Semiannual Operating Report, July to December 1974, Docket No.

50295.

13. Commonwealth Edison Company, Zion Station, Units 1 and 2, Semiannual Report on Radioactive Waste, Environmental Monitoring and Occupational Personnel Exposure, July to December 1975.

Docket No. 50295, February 1976.

14. The occupational exposure sections of the Annual Operating Reports as submitted to the NRC for the following stations:

Oconee Units 1, 2, and 3; Point Beach Units 1 and 2; Prairie Island Units 1 and 2; Surry Units 1 and 2; Turkey Point Units 3 and 4; Zion Units 1 and 2.

15. NUREG-0463, "Occupational Radiation Exposure, Tenth Annual Report, 1977," USNRC, October 1978.

12.4-4

B/B-UFSAR TABLE 12.4-1 PERSONNEL EXPOSURE DATA FOR VARIOUS OPERATING PWRs (1,15)*

MEASURABLY TOTAL AVERAGE MAN-REM MEGAWATT- EXPOSED ANNUAL EXPOSURE PER MEGA-STATION YEAR YEAR PERSONNEL MAN-REM (rem/person) WATT-YEAR Arkansas-1 75 588 147 21 0.14 0.04 850 MWe 76 463 476 289 0.61 0.62 77 610 601 256 0.43 0.42 Beaver Valley 77 328 331 87 0.26 0.27 852 MWe Calvert Cliffs-1 76 751 507 74 0.15 0.10 845 MWe 77 557 2265 547 0.24 0.98 D.C. Cook-1 76 805 395 116 0.29 0.14 1054 MWe 77 546 802 300 0.31 0.55 Fort Calhoun 75 252 469 294 0.63 1.17 457 MWe 76 265 516 313 0.61 1.18 77 334 535 297 0.56 0.89 Ginna 70 268.5 170 207 1.21 0.77 490 MWe 71 327.8 340 430 1.26 1.31 72 295.6 677 1032 1.52 3.49 73 409.5 319 224 0.70 0.55 74 253.7 884 1225 1.38 4.82 75 365 685 538 0.78 1.47 76 248 758 636 0.84 2.56 77 346 530 401 0.76 1.16 Haddam Neck 69 397.6 138 106 0.77 0.27 575 MWe 70 424.7 734 689 0.94 1.62 71 502 289 342 1.18 0.68 12.4-5

B/B-UFSAR TABLE 12.4-1 (Cont'd)

MEASURABLY TOTAL AVERAGE MAN-REM MEGAWATT- EXPOSED ANNUAL EXPOSURE PER MEGA-STATION YEAR YEAR PERSONNEL MAN-REM (rem/person) WATT-YEAR 72 515.6 355 325 0.91 0.63 73 293 951 697 0.73 2.38 74 519 550 201 0.37 0.39 75 494 795 703 0.88 1.42 76 482 644 449 0.70 0.93 77 458 894 642 0.72 1.40 Kewaunee 75 401.9 104 28 0.27 0.07 535 MWe 76 405 381 270 0.71 0.67 77 405 312 140 0.45 0.35 Maine Yankee 73 408.7 782 147 0.14 0.36 790 MWe 74 432.6 619 420 0.68 0.97 75 542.9 440 319 0.73 0.59 76 710 244 85 0.35 0.12 77 587 508 245 0.48 0.42 Oconee 1, 2, 3 74 724 844 517 0.61 0.71 886 MWe x 3 75 1084 829 497 0.60 0.46 76 1557 1215 1026 0.84 0.66 77 1485 1595 1329 0.83 0.90 Point Beach 1, 2 73 693.7 501 588 1.17 0.85 497 MWe x 2 74 760 400 295 0.74 0.39 75 801 339 459 1.35 0.57 76 855 313 370 1.18 0.43 77 834 417 430 1.03 0.52 12.4-6

B/B-UFSAR TABLE 12.4-1 (Cont'd)

MEASURABLY TOTAL AVERAGE MAN-REM MEGAWATT- EXPOSED ANNUAL EXPOSURE PER MEGA-STATION YEAR YEAR PERSONNEL MAN-REM (rem/person) WATT-YEAR Prairie Island 1, 2 74 181.9 150 18 0.12 0.10 530 MWe x 2 75 836 477 123 0.26 0.15 76 723 818 447 0.55 0.62 77 867 718 300 0.42 0.35 Robinson 71 295.3 283 364 1.28 1.23 700 MWe 72 580 245 215 0.87 0.37 73 455 831 695 0.83 1.53 74 577 853 672 0.78 1.16 75 501.8 849 1142 1.35 2.28 76 584 597 715 1.20 1.22 77 493 634 455 0.72 0.92 San Onofre 1 69 289.8 123 42 0.34 0.14 436 MWe 70 365.9 251 155 0.61 0.42 71 362 121 50 0.41 0.14 72 372 326 256 0.78 0.69 73 273.7 878 329 0.37 1.20 74 377.8 219 71 0.32 0.19 75 389 424 292 0.69 0.75 76 297 1330 880 0.66 2.96 77 266 985 847 0.86 3.18 Surry 1, 2 73 714 936 152 0.16 0.21 823 MWe x 2 74 718 1715 884 0.52 1.23 75 1079 1948 1649 0.85 1.53 76 928 2753 3165 1.15 3.41 77 1082 1860 2307 1.24 2.13 12.4-7

B/B-UFSAR TABLE 12.4-1 (Cont'd)

MEASURABLY TOTAL AVERAGE MAN-REM MEGAWATT- EXPOSED ANNUAL EXPOSURE PER MEGA-STATION YEAR YEAR PERSONNEL MAN-REM (rem/person) WATT-YEAR Three Mile 75 675.9 ~168 ~83 ~0.49 ~0.1 Island 1 76 529 819 286 0.35 0.54 819 MWe 77 624 1122 360 0.32 0.58 Trojan 1 77 741 591 174 0.29 0.24 1130 MWe Turkey Point 3, 4 73 402 444 78 0.18 0.19 745 MWe x 2 74 953 794 454 0.57 0.48 75 1003.7 1176 876 0.74 0.87 76 972 1647 1184 0.72 1.22 77 928 1319 1036 0.79 1.12 Zion 1, 2 74 424 306 56 0.18 0.13 1040 MWe x 2 75 1181 436 127 0.29 0.11 76 1132 774 571 0.74 0.50 77 1291 784 1004 1.28 0.78 78 1578 1104 952 0.86 0.60 Average** 0.72 0.84

The number of personnel includes station personnel, contractors, and temporary workers. Generally, only the number of individuals with exposures greater than 100 mrems is reported.

- Averages include values corresponding to the first year in which the power generated was 55% or more of the rated output and values corresponding to all subsequent years.

12.4-8

B/B-UFSAR TABLE 12.4-2 PERSONNEL EXPOSURE DATA FOR MULTIPLE-UNIT OPERATING PWRs (1,15)

TOTAL ANNUAL AVERAGE EXPOSURE (REM/PERSON)

STATION YEAR MAN-REM TOTAL CONTRACTOR UTILITY Oconee 1, 2, 3 74 517 0.61 0.57 0.63 886 MWe x 3 75 457 0.84 0.74 0.87 76 987 1.07 0.84 1.15 Point Beach 1, 2 72 580 497 MWe x 2 73 570 0.78 74 295 0.74 75 456 1.3 76 362 1.40 0.89 1.84 Prairie Island 1, 2 74 18 0.12 0.09 0.14 530 MWe x 2 75 123 0.26 0.22(3,4) 0.23(3,4) 76 424 0.83 Surry 1, 2 73 152 0.16 823 MWe x 2 74 884 0.52 75 1549 1.91 76 3060 1.57 1.34 2.07 Turkey Point 3, 4 73 78 0.18 745 MWe x 2 74 454 0.57 75 875 0.74 76 1408 1.21 1.43 0.86 12.4-9

B/B-UFSAR TABLE 12.4-2 (Cont'd)

TOTAL ANNUAL AVERAGE EXPOSURE (REM/PERSON)

STATION YEAR MAN-REM TOTAL CONTRACTOR UTILITY Zion 1, 2 74 33 0.18 0.15 0.20 1050 MWe x 2 75 118 0.08 0.05 0.15 76(5) 525 0.31 0.23 0.46 Average (See Note) 543 0.78 -- --

Byron/Braidwood original Estimated Annual Man-rem 800 (1) Based on total annual man-rem average 543 (2) Based on 250 station employees) at 0.078 rem/person 800 775 contract workers)

NOTES:

Averages include values corresponding to the first year in which the power generated was 55% or more of the rated output, and values corresponding to all subsequent years. (See Table 12.4-1)

Data for Surry is not included in the averages, since the steam generator tube failures which resulted in high man-rem exposures at Surry are not expected to occur at Byron/Braidwood, which has a different steam generator design and all-volatile chemistry for feedwater conditioning.

The predicted occupational exposure at Zion for 1977 is 700-750 man-rem.

12.4-10 REVISION 1 - DECEMBER 1989

B/B-UFSAR TABLE 12.4-3 REPORTED PERSONNEL EXPOSURE BY WORK FUNCTION FOR SEVERAL OPERATING PWRs (5-14)

ROUTINE MAINTEN- RADWASTE ROUTINE OPER- ANCE AND PROCESSING SPECIAL ATIONS AND INSERVICE AND MAIN-STATION YEAR SURVEILLANCE INSPECTION REFUELING HANDLING OTHER TENANCE Ginna 1975 27 192 61 19 12 180 Haddam Neck 1975 30 185 64 7 1 381 Maine Yankee 1975 25 105 138 27 NR 0 Average (single-unit plants) 27 161 88 18 - 187 Oconee 1, 2, 3 1976 63 180 138 30 NR 575 Point Beach 1, 2 1976 56 148 125 26 NR 6 Prairie Island 1,2 1976 64 58 32 8 NR 262 Surry 1, 2 1974 46* 373 43 NR NR 127 1975 39 104 47 90 NR 160 1976 429 1210 133 NR NR 1287 Turkey Point 3,4 1976 111 977 9 24 NR 293 Zion 1,2 1975 16 190 3 8 NR NR 1976 59 162 13 14 NR NR 1977 42 334 18 26 NR 11 1978 141 299 22 24 NR 75 12.4-11

B/B-UFSAR TABLE 12.4-3 (Cont'd)

ROUTINE MAINTEN- RADWASTE ROUTINE OPER- ANCE AND PROCESSING SPECIAL ATIONS AND INSERVICE AND MAIN-STATION YEAR SURVEILLANCE INSPECTION REFUELING HANDLING OTHER TENANCE Average (multiple-unit plants) 66 290 49 19 - 247 (See Notes)

Byron/Braidwood Estimated Annual Man-rem Routine Maintenance and Surveillance 65 Routine Maintenance and Inservice Inspection 300 Refueling 65 Radwaste Processing and Handling 20 Other 50 Special Maintenance 300 TOTAL 800 12.4-12

B/B-UFSAR TABLE 12.4-3 (Cont'd)

NOTES:

Exposures given were reported as the sum of individual exposures greater than 500 mrem, except for Zion.

Where the breakdown in the original report was more detailed, categories have been condensed as necessary to obtain the categories given here.

The category "other" includes training, miscellaneous, security, consultants, etc.

Where data was incomplete for one-half of the year, the data was prorated from the other complete half of the year, except for refueling.

NR means "not reported," that is, no data for this or any similar category was reported.

"Normal surveillance" only was reported.

Data for Surry is not included in the averages, since the steam generator tube failures which resulted in high man-rem exposures at Surry are not expected to occur at Byron/Braidwood, which has a different steam generator design and an all-volatile chemistry for feedwater conditioning.

Estimates are conservative to account for exposures less than 100 mrem which are not generally included in reports of occupational exposure and thus are not included in the averages.

12.4-13

B/B-UFSAR TABLE 12.4-4 ANNUAL THYROID DOSES RESULTING FROM CALCULATED DESIGN-BASIS AIRBORNE CONCENTRATIONS IN REMS/YR AUXILIARY* CONTAINMENT** RADWASTE***

ISOTOPE BUILDING BUILDING BUILDING I-131 6.7 x 10-1 1.67 1.8 x 10-3 I-132 9.0 x 10-3 1.65 x 10-2 negligible I-133 2.7 x 10-1 5.0 x 10-1 negligible I-134 3.1 x 10-3 8.2 x 10-4 negligible I-135 4.7 x 10-2 3.3 x 10-2 negligible The above dose rates are based on 13.3 hr/wk exposure of personnel in the auxiliary building, of which 50%

is spent in clean areas, 35% in general areas with potential airborne, 10% in pump room and valve aisle, and 5% in radiation areas.

Reactor building dose rates are based on 13.3 hr/wk of which 1% is spent in the containment.

Radwaste building dose rates are based on 13.3 hr/wk with 5% occupation time.

12.4-14

B/B-UFSAR TABLE 12.4-5 ESTIMATED FIFTH YEAR RADIATION DOSE FOR B/B COMPARED WITH 1976 AND 1977 OPERATING DATA*

2 UNIT AVERAGE PER UNIT* STATIONS* 1976-1978 B/B 1&2 WORK FUNCTION 1976 1977 1976 1977 ZION 1&2 (ESTIMATED)

Routine Operations 38 36 66 73 81 70 and Surveillance Routine Inspection 164 150 284 303 279 260 and Maintenance Refueling 29 22 51 46 19 20 Radwaste 19 20 32 41 21 10 Special Maintenance 123 110 210 230 420 230 TOTAL 373 338 643 693 820 590 Power Rating (761 Mwe) (1380 Mwe) (2100 Mwe) (2300 Mwe)

PWR operating data taken from NUREG-0463, Table 7 and Appendix A minus the Surry 1&2 data. The units are man-rems unless designated otherwise.

12.4-15

B/B-UFSAR 12.5 HEALTH PHYSICS PROGRAM 12.5.1 Organization The administrative organization of the health physics program and personnel responsibilities are referenced in Subsections 12.1.1.1 and 12.1.1.2.

The experience and qualification of all station personnel are given in station procedures.

12.5.2 Equipment, Instrumentation, and Facilities Table 12.5-1 lists the normal storage location of respiratory protective equipment, protective clothing, and portable and laboratory technical equipment and instrumentation.

Respiratory protective equipment is used to limit intakes of airborne radioactive material when engineering controls are not feasible and when consistent with the principle of minimizing total effective dose equivalent. The following types of respirators are among those available for use: air purifying full mask respirators, air line full mask respirators, air line airborne hood respirators, and positive pressure self-contained breathing apparatus (SCBA). At Byron, in addition to the equipment listed in Table 12.5-1, a reserve of emergency breathing air is maintained for control room personnel. At Braidwood, in addition to the equipment listed in Table 12.5-1, emergency breathing air for control room personnel is provided by additional SCBAs with bottled air available for backup.

Typical estimates for and the quantity, sensitivity, range, and frequency and methods of calibration for health physics instrumentation and technical equipment are specified in Table 12.5-2.

Table 12.5-2 shows portable radiation monitoring instrumentation capable of measuring exposure rates up to 10,000 R/hr. Such instrumentation would be used under accident conditions in areas where it is impractical to have installed stationary monitors.

Since source calibration of high range instrumentation is impractical on the upper scales, only electronic calibrations will be performed for the upper scales/decades of high range exposure rate instrumentation.

Health physics and radiochemist facilities are described in Table 12.5-3.

12.5.3 Procedures The health physics procedures have been developed to implement Exelon Generation Company's commitment to "As Low As Reasonably Achievable" (ALARA) as stated in Subsection 12.1.1.

12.5-1 REVISION 8 - DECEMBER 2000

B/B-UFSAR 12.5.3.1 Administrative Program Strict administrative control of radiation exposure includes those methods described in Subsections 12.5.3.2, 12.5.3.3, and 12.5.3.5. Other administrative controls used include locked high radiation areas, radiation work permits, timekeeping of personnel in high radiation areas, and security measures including escorts for visitors within the plant security area.

12.5.3.2 Personnel External Exposure Program The personnel external exposure program consists of multiple methods of reviewing external radiation levels and controls within the plant. These provide plant and personnel status information required to maintain an ALARA program (Subsection 12.1.1).

Area radiation monitors (ARMs) are located throughout the plant and provide general area indication of gamma radiation levels.

These levels are continuously monitored and are alarmed in the control room. Some monitors also have local indication and alarm at certain in-plant locations. Besides surveillance by control room operators, these levels are periodically reviewed by a health physicist to note unusual trends. Process radiation monitors with control room indication and alarms also provide for immediate recognition of significant increases in in-plant dose rate levels.

Routine beta-gamma dose rate surveys are made of general access areas of the plant. This provides detailed dose rate information for normal in-plant exposure evaluation. The survey sheets are reviewed to note unusual trends and for determination of additional controls that may be required due to new or increased radiation dose rates.

Special beta-gamma dose rate surveys are made on an as needed basis for jobs that take place in normally inaccessible (i.e.,

high radiation) areas. These areas are not normally surveyed on a routine basis due to the required dose commitment being inconsistent with the ALARA program. Continuous or intermittent surveys are provided on an as needed basis as determined by radiation protection for radiation work permits (Subsection 12.5.3.1).

Personnel entering radiologically posted areas onsite are required to wear appropriate dosimetry. This dosimetry consists of a primary dosimeter of legal record and electronic self-reading dosimeters. Daily, the self-reading dosimeter readings (or equivalent) and timekeeping results (if applicable) are normally recorded, and are routinely reviewed by radiation protection management and by management in the individual's work group, if applicable. The primary dosimeters of legal record are changed routinely. Badge results are reviewed and are entered in the Exelon Generation Company computerized 12.5-2 REVISION 14 - DECEMBER 2012

B/B-UFSAR radiation exposure records system. These official and permanent records furnish the exposure data for the administrative control of radiation exposure. Required reports are made by radiation protection management through the use of this records system.

General area neutron dose rate measurements are made during startup after initial fuel loading and following refueling outages to verify neutron dose rates. Special neutron surveys and use of neutron dosimeters are provided when entrance is made into neutron areas when required by 10 CFR 20.

Radioactive materials and special nuclear materials are handled and stored under the direction of personnel as specified in Subsection 12.1.1.2.

12.5.3.3 Personnel Internal Exposure Program The personnel internal exposure program consists of multiple methods of reviewing airborne radioactivity concentrations and controls within the plant. These provide plant and personnel status information required to maintain an ALARA program (Subsection 12.1.1).

The Station vent stack monitors (one for each of the two vent stacks) have detectors for air particulate, gas (low and high range), iodine, and background subtraction. In addition to surveillance by control room operators, monitor levels are periodically reviewed by radiation protection personnel to note unusual trends.

Continuous air monitors also monitor auxiliary building ventilation exhausts, containment purge systems, and the radwaste building ventilation exhausts. These are used to measure, indicate, and record levels of airborne radioactivity in air exhausted from plant areas and as trending devices by radiation protection personnel.

Portable grab samples are normally taken in accessible areas of the plant on a periodic basis. Special samples are taken as required by radiation protection personnel prior to issuing Radiation Work Permits and before other jobs as necessary. These air sample results are reviewed by radiation protection personnel and are used to determine respiratory protective equipment requirements in accordance with Station radiation protection procedures.

Whole body counts are performed for plant personnel with a frequency as specified in the station radiation protection procedures.

12.5-3 REVISION 13 - DECEMBER 2010

B/B-UFSAR All personnel (permanent and temporary) are normally requested to have a whole body count or whole body screening before termination if they have worked in airborne radioactivity or with radioactive materials unless specifically exempted by radiation protection 12.5-3a REVISION 8 - DECEMBER 2000

B/B-UFSAR management. Other bioassay techniques may be substituted, such as urinalysis and fecal analysis. A personnel bioassay program is administered by a health physicist. Bioassay (in vivo measurement and/or measurement of radioactive material in excreta) are conducted as necessary to aid in determining the extent of an individual's internal exposure to concentration of radioactive material. The need for and frequency of bioassay are determined by the duration that a person works with radioactive materials or in an airborne radioactivity area. Specific frequencies are determined and controlled by procedures. All bioassay results are recorded as required.

The Byron/Braidwood bioassay program is implemented in compliance with Revision 1 of Regulatory Guide 8.9, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program."

12.5-4 REVISION 8 - DECEMBER 2000

B/B-UFSAR 12.5.3.4 Contamination Control Program The contamination control program consists of multiple methods of controlling the spread of contamination to personnel and equipment within the plant. Routine smear surveys are periodically made of normally accessible areas of the plant and are recorded on survey sheets. These results are reviewed by a radiation protection supervisor or a health physicist. Special smear surveys are performed on an as needed basis for radiation work permits and for unconditional release of equipment, tools, and materials being removed from radiologically posted areas.

Items which are contaminated are required to be decontaminated to within release limits or packaged and tagged in accordance with the station radiation protection procedures.

Workers in contaminated areas are required to be monitored for contamination prior to leaving the contamination control point for the areas. Additionally, portal-type monitors are utilized to monitor individuals leaving the radiologically posted area (RPA) via the main access area and again when leaving the site (in the security gatehouse). Actual instrumentation used for the contamination surveys is determined by station management.

12.5.3.5 Training Program The radiation protection training programs are described in Section 13.2. This program covers the following:

a. general employee health physics,

b. general employee respiratory protection,

c. contractor health physics,

d. contractor respiratory protection,

e. general employee retraining,

f. Radiation Protection Technician training, and

g. Radiation Protection Technician retraining.

All personnel must understand how radiation protection relates to their jobs and have reasonable opportunities to discuss radiation protection safety with a member of the Radiation Protection department whenever the need arises. Plant personnel are made aware of Exelon Generation Company commitment to keep occupational radiation exposure as low as reasonably achievable (Subsection 12.1.1). A minimum goal of this program is that workers shall be sufficiently familiar with this commitment that they can explain what the management commitment is, what "As Low As Reasonably Achievable" means, why it is recommended, and how they have been advised to implement it on their jobs.

12.5-5 REVISION 8 - DECEMBER 2000

B/B-UFSAR Qualifications of personnel, including training requirements for radiation protection personnel are described in Subsection 12.5.1.

12.5-5a REVISION 8 - DECEMBER 2000

B/B-UFSAR TABLE 12.5-1 STORAGE LOCATION OF EQUIPMENT EQUIPMENT NORMAL STORAGE LOCATION Self-Contained Breathing Apparatus Control Room, Technical Support Center, Operational (Pressure Demand) Support Center Full Face Masks (Air Purifying) Auxiliary Building - Mask Storage Area Full Face Masks (Airline)

Hoods (Airline)

Protective Clothing Auxiliary Building

- Air Ionization Chambers Auxiliary Building - Calibration Facility G-M Survey Instruments Neutron Detector Chemical Analysis Equipment Hot Laboratory, Cold Laboratory 12.5-6 REVISION 8 - DECEMBER 2000

B/B-UFSAR TABLE 12.5-2 HEALTH PHYSICS EQUIPMENT ESTIMATED TYPE DETECTOR/MONITOR* NUMBER SENSITIVITY RANGE FREQUENCY CALIBRATION METHOD Gamma Ray Counting System 2 Variable Variable Per CY-AA approved Standard Reference procedure Materials Alpha/Beta Counting 2 Variable Variable Per CY-AA approved Standard Reference System procedure Materials Air Ion Chamber Exposure 30 Variable Variable Annual Standard Reference Rate Meter Materials GM Survey Count 15 Variable Variable Annual Standard Reference Rate Instrument Materials Alpha 2 Variable 0-100K cpm Annual Standard Reference Scintillator Probe Materials High Range, Exposure 5 Variable 0-10,000 R/hr Annual Standard Reference Rate Materials Neutron Detector 2 Variable 0-5 Rem/hr Annual Standard Reference Materials/Mini Pulser Air Sampler 10 N/A Variable Annual Air Flow Calibrator Portable Continuous 5 Variable 0-50K cpm Annual Manometer, Standard Air Monitor 0-10 cfm Reference Materials

The instrument/equipment list is intended to be typical of in-service instrumentation.

12.5-7 REVISION 14 - DECEMBER 2012

B/B-UFSAR THIS PAGE DELETED INTENTIONALLY.

12.5-8 REVISION 8 - DECEMBER 2000

B/B-UFSAR TABLE 12.5-3 HEALTH PHYSICS AND RADIOCHEMICAL FACILITIES NAME LOCATION PRIMARY FUNCTION Calibration Facility Auxiliary Building Calibration of Gamma Dose Rate Instruments and Storage of Survey Instruments Hot Laboratory Auxiliary Building Chemical Analysis and Radiochemical Separations Cold Laboratory Auxiliary Building Chemical Analysis Supply Room Auxiliary Building Storage of Chemicals, Glassware, and (Braidwood only) Laboratory Equipment Counting Room Auxiliary Building Radioactivity and Radiological Determination of Samples Laundry Room* Auxiliary Building Storage of Protective Radiological (Byron) Clothing Laundry Room* Auxiliary Building Store equipment and supplies, sort low (Braidwood) level radioactive trash, and launder personal clothing Mask Cleaning Facility Auxiliary Building Cleaning, Inspection, and Storage of Respiratory Equipment Health Physics Offices Turbine Building Administration/Offices Area/Service Building

An offsite vendor is utilized to clean potentially contaminated protective clothing.

12.5-9 REVISION 10 - DECEMBER 2004

B/B-UFSAR Figure 12.2-1 has been deleted intentionally.

REVISION 13 - DECEMBER 2010

BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 12.3-1 SKETCH OF A SIMPLE LABYRINTH ENTRANCE

I+w.

~~

~ 1-1"" ...

BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 12.3-2 SKETCH OF A DOUBLE LABYRINTH ENTRANCE

VALVE AISLE

<,. ... '" ../'< ,....,,,,<"7 L. ",v,.." .. " <'If DOOR VALVE OPERATING AREA LAYOUT FOR WALK-IN VALVE AISLE (PLAN VIEW)

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r ,. ....,;,........~--"""'l REACH RODS ARE ROUTED:

"7 CD TO PERMIT "BONNET-UP"

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BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 12.3-3 TYPICAL WALK*IN VALVE AISLE

Shield Hatch CONCRETE OR STEEL SHIELD PLUG WHICH CAN BE REMOVED FOR INSERTION OF RADIATION DETECTOR BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 12.3-4 SKETCH OF RADIATION DETECTOR PROBE ACCESS HOLE IN SHIELD HATCH FOR FILTER OR DEMINERAUZER

B/B-UFSAR Figure 12.3-5 through 12.3-26a have been deleted intentionally.

REVISION 9 - DECEMBER 2002

Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 Security - Related Information Figure Withheld Under 10 CFR 2.390 A =-e.'"

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UNIT 2 15 OPPOSITE HAND BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 12.3A-4 REMOVABLE BLOCK WALL PLAN

Security - Related Information Figure Withheld Under 10 CFR 2.390

ML19170A387

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