ML13004A059

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Units 1 and 2, Updated Final Safety Analysis Report, Revision 14, Chapter 5.0 - Reactor Coolant System and Connected Systems
ML13004A059
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 12/14/2012
From:
Exelon Generation Co
To:
Office of Nuclear Reactor Regulation, Office of Nuclear Material Safety and Safeguards
References
RS-12-221
Download: ML13004A059 (251)
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Text

B/B-UFSAR 5.0-i CHAPTER 5.0 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS TABLE OF CONTENTS PAGE 5.0 REACTOR COOLANT SYSTEM AND CON NECTED SYSTEMS 5.1-1 5.1

SUMMARY

DESCRIPTION 5.1-1 5.1.1 Schematic Flow Diagram 5.1-6 5.1.2 Piping and Instrumentation Diagrams 5.1-6 5.1.3 Elevation Drawings 5.1-6

5.2 INTEGRITY

OF REACTOR COOLANT PRESSURE BOUNDARY 5.2-1 5.2.1 Compliance with Codes and Code Cases 5.2-1 5.2.1.1 Compliance with 10 CFR 50.55a 5.2-1 5.2.1.2 Applicable Code Cases 5.2-2

5.2.2 Overpressurization

Protection 5.2-3 5.2.2.1 Design Bases 5.2-3 5.2.2.2 Design Evaluation 5.2-4 5.2.2.3 Piping and Instrumentation Diagrams 5.2-5 5.2.2.4 Equipment and Co mponent Description 5.2-6 5.2.2.5 Mounting of Pressure-Relief Devices 5.2-6 5.2.2.5.1 Design and Installation Details 5.2-6 5.2.2.5.1.1 Pressurizer Sa fety Valves and Power- Operated Relief Valves 5.2-6 5.2.2.5.1.2 Main Steam S afety Valves and Power- Operated Relief Valves 5.2-6 5.2.2.5.2 Design Bases for Assumed Loads 5.2-6 5.2.2.5.3 Maximum Stress 5.2-7 5.2.2.6 Applicable Codes and Classifications 5.2-7 5.2.2.7 Material Specifications 5.2-8 5.2.2.8 Process Instrumentation 5.2-8 5.2.2.9 System R eliability 5.2-8 5.2.2.10 Testing and Inspection 5.2-8 5.2.2.11 RCS Pressure Control During Low Temperature Operation 5.2-8 5.2.2.11.1 System Operation 5.2-9 5.2.2.11.2 Evaluation of Low Temperature Overpressure Transients 5.2-9 5.2.2.11.3 Procedures 5.2-13 5.2.3 Reactor Coolant Pressure Boundary Materials 5.2-15 5.2.3.1 Material Specifications 5.2-15 5.2.3.2 Capability with Reactor Coolant 5.2-16 5.2.3.2.1 Chemistry of Reactor Coolant 5.2-16 5.2.3.2.2 Compatibility of Construction Materials with Reactor Coolant 5.2-18 5.2.3.2.3 Compatibility with External Insulation and Environmental Atmosphere 5.2-18

B/B-UFSAR 5.0-ii REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS (Cont'd) 5.2.3.3 Fabrication and Processing of Ferritic Materials 5.2-19 5.2.3.3.1 Fracture Toughness 5.2-19 5.2.3.3.2 Control of Welding 5.2-19 5.2.3.4 Fabrication and Proc essing of Austenitic Stainless Steel 5.2-20 5.2.3.4.1 Cleaning and C ontamination Protection Procedures 5.2-20 5.2.3.4.2 Solution Heat Tr eatment Requirements 5.2-21 5.2.3.4.3 Material Inspection Program 5.2-21 5.2.3.4.4 Prevention of Intergranular Attack of Unstabilized Austenitic Stainless Steels 5.2-22 5.2.3.4.5 Retesting Unstab ilized Austenitic Stainless Steels Exposed to Sensit ization Temperatures 5.2-25 5.2.3.4.6 Control of Welding 5.2-25 5.2.4 Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.2-27 5.2.5 Detection of Leakage Through Coolant Pressure Boundary 5.2-29 5.2.5.1 Reactor Cavity a nd Containment Floor Drain Sumps 5.2-30 5.2.5.2 Containment Radiation Monitoring 5.2-31 5.2.5.2.1 Radiation Monitor Se nsitivity/Response Time 5.2-32 5.2.5.2.2 Leak Before Break Considerations 5.2-32 5.2.5.3 Containment Atmosphere Monitoring 5.2-32 5.2.5.4 Intersystem Leakage 5.2-32a 5.2.5.5 Intersystem Leakage Monitoring 5.2-34 5.2.5.6 Limiting Conditions for Operation 5.2-35 5.2.5.7 Intersystem Leakage Testing 5.2.35 5.2.5.8 Reactor Vess el Flange Leakage Monitoring 5.2-35 5.2.5.9 Calibration and Oper ability Tests During Plant Operation 5.2-36 5.2.6 References 5.2-37

5.3 REACTOR

VESSEL 5.3-1 5.3.1 Reactor Vessel Materials 5.3-1 5.3.1.1 Material Specifications 5.3-1 5.3.1.2 Special Processes Us ed for Manufacturing and Fabrication 5.3-1 5.3.1.3 Special Methods for Nondestructive Examination 5.3-2 5.3.1.3.1 Ultrasonic Examination 5.3-2 5.3.1.3.2 Penetrant Examinations 5.3-2 5.3.1.3.3 Magnetic Particle Examination 5.3-2 5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels 5.3-3 5.3.1.5 Fracture Toughness 5.3-3 5.3.1.5.1 Pressurized Thermal Shock Evaluation 5.3-4 5.3.1.6 Material Surveillance 5.3-4 5.3.1.6.1 Measurement of I ntegrated Fast Neutron (E>1.0MeV) Flux at the Irradiation Samples 5.3-6 5.3.1.6.1.1 Determ ination of Sensor Reaction Rates 5.3.1.6.1.2 Corrections to Reaction Rate Data 5.3.1.6.1.3 Least Square s Adjustment Procedure B/B-UFSAR 5.0-iia REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS (Cont'd) 5.3.1.6.2 Calculation of I ntegrated Fast Neutron (E>1.0MeV) Flux at the Irradiation Samples 5.3-8 5.3.1.6.2.1 Reference Forward Calculation 5.3.1.6.2.2 Cycle Specif ic Adjoint Calculations 5.3.1.7 Reactor Vessel Fasteners 5.3-8

B/B-UFSAR 5.0-iii REVISION 7 - DECEMBER 1998 TABLE OF CONTENTS (Cont'd) 5.3.2 Pressure-Temperature Limits 5.3-9 5.3.2.1 Limit Curves 5.3-9 5.3.2.2 Operating Procedures 5.3-10 5.3.3 Reactor Vessel Integrity 5.3-10 5.3.3.1 Design 5.3-10 5.3.3.2 Materials of Construction 5.3-11 5.3.3.3 Fabrication Methods 5.3-11 5.3.3.4 Inspection Requirements 5.3-11 5.3.3.5 Shipment and Installation 5.3-12 5.3.3.6 Operating Conditions 5.3-12 5.3.3.7 Inservice Surveillance 5.3-12 5.3.4 References 5.3-14

5.4 COMPONENT

AN D SUBSYSTEM DESIGN 5.4-1 5.4.1 Reactor Coolant Pumps 5.4-1 5.4.1.1 General 5.4-1 5.4.1.2 Design D escription 5.4-1 5.4.1.3 Design Evaluation 5.4-3 5.4.1.3.1 Pump Performance 5.4-3 5.4.1.3.2 Coastdown Capability 5.4-4 5.4.1.3.3 Bearing Integrity 5.4-4 5.4.1.3.4 Locked Rotor or Loss of CCW 5.4-5 5.4.1.3.5 Critical Speed 5.4-7 5.4.1.3.6 Missile Generation 5.4-7 5.4.1.3.7 Pump Cavitation 5.4-7 5.4.1.3.8 Pump Overspeed Considerations 5.4-8 5.4.1.3.9 Antireverse Rotation Device 5.4-8 5.4.1.3.10 Shaft Seal Leakage 5.4-8 5.4.1.3.11 Seal Discharge Piping 5.4-13 5.4.1.4 Tests and Inspections 5.4-13 5.4.1.5 Pump Flywheels 5.4-14 5.4.1.5.1 Design Bases 5.4-14 5.4.1.5.2 Fabrication and Inspection 5.4-14 5.4.1.5.3 Material Acceptance Criteria 5.4-15 5.4.2 Steam Generators 5.4-15 5.4.2.1 Steam Generator Materials 5.4-15 5.4.2.1.1 Selection and Fabrication of Materials 5.4-15 5.4.2.1.2 Steam Genera tor Design Effects on Materials 5.4-16 5.4.2.1.3 Compatibility of Steam Generator Tubing with Primary and Secondary Coolants 5.4-16 5.4.2.1.4 Cleanup of Secondary Side Materials 5.4-18 5.4.2.2 Steam Genera tor Inservice Inspection 5.4-18 5.4.2.3 Design Bases 5.4-18 5.4.2.4 Design D escription 5.4-20 5.4.2.5 Design Evaluation 5.4-20 5.4.2.5.1 Forced Convection 5.4-20 5.4.2.5.2 Natural Circulation Flow 5.4-21 5.4.2.5.3 Mechanical and Flow-Induced Vibration Under Normal Operation 5.4-21 5.4.2.5.4 Allowable Tube Wall Thinning Under Accident Conditions 5.4-23 5.4.2.6 Tests and Inspection 5.4-24

B/B-UFSAR 5.0-iv REVISION 3 - DECEMBER 1991 TABLE OF CONTENTS (Cont'd)

5.4.3 Reactor

Coolant Piping 5.4-25 5.4.3.1 Design Bases 5.4-25 5.4.3.2 Design D escription 5.4-26 5.4.3.3 Design Evaluation 5.4-29 5.4.3.3.1 Material Corrosion/Erosion Evaluation 5.4-29 5.4.3.3.2 Sensitized Stainless Steel 5.4-30 5.4.3.3.3 Contaminant Control 5.4-30 5.4.3.4 Tests and Inspections 5.4-30 5.4.4 Main Steamline Flow Restrictions 5.4-31 5.4.4.1 Design Bases 5.4-31 5.4.4.2 Design D escription 5.4-31 5.4.4.3 Design Evaluation 5.4-31 5.4.4.4 Tests and Inspections 5.4-31 5.4.5 Main Steamline Isolati on System (BWRs Only) 5.4-31 5.4.6 Reactor Core Iso lation Cooling System (BWRs Only) 5.4-32 5.4.7 Residual Heat Removal System 5.4-32 5.4.7.1 Design Bases 5.4-32 5.4.7.2 System Design 5.4-34 5.4.7.2.1 Schematic Pipi ng and Instrumentation Diagrams 5.4-34 5.4.7.2.2 Equipment and Co mponent Descriptions 5.4-37 5.4.7.2.3 Control 5.4-38 5.4.7.2.4 Applicable Codes and Classifications 5.4-40 5.4.7.2.5 System Reliability Considerations 5.4-40 5.4.7.2.6 Manual Actions 5.4-46 5.4.7.2.7 System Operation 5.4-47 5.4.7.3 Performance Evaluation 5.4-56 5.4.7.4 Tests and Inspections 5.4-57 5.4.8 Reactor Water Cleanup System (BWRs Only) 5.4-59 5.4.9 Main Steam Line and Feedwater Piping 5.4-59 5.4.10 Pressurizer 5.4-60 5.4.10.1 Design Bases 5.4-60 5.4.10.1.1 Pressurizer Surge Line 5.4-60 5.4.10.1.2 Pressurizer Volume 5.4-60 5.4.10.2 Design Description 5.4-61 5.4.10.2.1 Pressurizer Surge Line 5.4-61 5.4.10.2.2 Pressurizer Vessel 5.4-61 5.4.10.3 Design Evaluation 5.4-62 5.4.10.3.1 System Pressure 5.4-62 5.4.10.3.2 Pressurizer Performance 5.4-63 5.4.10.3.3 Pressure Setpoints 5.4-63 5.4.10.3.4 Pressurizer Spray 5.4-63 5.4.10.3.5 Pressurizer Design Analysis 5.4-64 5.4.10.4 Tests and Inspections 5.4-65 5.4.11 Pressurizer Relief Discharge System 5.4-66 5.4.11.1 Design Bases 5.4-66 5.4.11.2 System Description 5.4-66 5.4.11.3 Safety Evaluation 5.4-67 5.4.11.4 Instrumentation Requirements 5.4-68 5.4.11.5 Inspection and T esting Requirements 5.4-68

B/B-UFSAR 5.0-v TABLE OF CONTENTS (Cont'd) 5.4.12 Valves 5.4-69 5.4.12.1 Design Bases 5.4-69 5.4.12.2 Design Description 5.4-69 5.4.12.3 Design Evaluations 5.4-70 5.4.12.4 Tests and Inspections 5.4-70 5.4.13 Safety and Relief Valves 5.4-70 5.4.13.1 Design Bases 5.4-70 5.4.13.2 Design Description 5.4-71 5.4.13.3 Design Evaluation 5.4-72 5.4.13.4 Tests and Inspections 5.4-72 5.4.14 Component Supports 5.4-73 5.4.14.1 Reactor Vessel Supports 5.4-73 5.4.14.2 Pressurizer Support 5.4-73 5.4.14.3 Steam Generator Support 5.4-73 5.4.14.4 Reactor Coolant Pump Support 5.4-73 5.4.14.5 Design Criteria for Component Supports 5.4-73 5.4.15 References 5.4-73

B/B-UFSAR 5.0-vi REVISION 9 - DECEMBER 2002 CHAPTER 5.0 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS LIST OF TABLES NUMBER TITLE PAGE 5.1-1 System Design and Operating Parameters 5.1-7 5.1-1a RCS Component Volume Data 5.1-10 5.2-1 Applicable Code Addenda for RCS Components 5.2-38 5.2-la ASME Code Cases Used on Class 1 Components 5.2-39 5.2-2 Class 1 Primary Components Material Specifications 5.2-40 5.2-3 Class 1 and 2 Auxiliary Components Material Specifications 5.2-43 5.2-4 Reactor Vessels Inte rnals for Emergency Core Cooling 5.2-45 5.3-1 Reactor Vessel Nonde structive Examination During Fabrication 5.3-15 5.3-2 Reactor Vessel Design Parameters 5.3-17 5.3-3a Byron Unit 1 Clo sure Heat Bolting Material Properties Closure Head Studs 5.3-18 5.3-3b Braidwood Unit 1 Closure Head Bolting Material Properties Closure Head Studs 5.3-19 5.3-4 (Deleted) 5.3-20 5.3-5 (Deleted) 5.3-20 5.3-6 (Deleted) 5.3-20 5.3-7 Byron Unit 1 Pressur ized Thermal Shock (PTS) Evaluation 5.3-21 5.3-8 Byron Unit 2 Pressur ized Thermal Shock (PTS) Evaluation 5.3-22 5.3-9 Braidwood Unit 1 Press urized Thermal Shock Evaluation 5.3-23 5.3-10 Braidwood Unit 2 Pressurized Thermal Shock Evaluation 5.3-24 5.4-1 Reactor Coolant Pump Design Parameters 5.4-74 5.4-2 Reactor Coolant Pump N DE During Fabrication 5.4-76 5.4-3 Steam Generator Design Data 5.4-77 5.4-4 Steam Generator NDE During Fabrication (Unit 1) 5.4-78 5.4-4a Steam Generator NDE During Fabrication (Unit 2) 5.4-79a 5.4-5 Reactor Coolant Piping Design Parameters 5.4-80 5.4-6 Reactor Coolant Piping NDE During Fabrication 5.4-81 5.4-7 Design Bases for Residual Heat Removal System Operation 5.4-82 5.4-8 Residual Heat Remova l System Component Data 5.4-83 5.4-9 Pressurizer Design Data 5.4-84 5.4-10 Reactor Coolant System Design Pressure Settings 5.4-85 5.4-11 Pressurizer Quality Assurance Program 5.4-86 5.4-12 Pressurizer Relief Tank Design Data 5.4-87 5.4-13 Relief Valve Discharge to the Pressurizer Relief Tank 5.4-88 5.4-14 Reactor Coolant System Valve Design Parameters 5.4-89

B/B-UFSAR 5.0-vii REVISION 9 - DECEMBER 2002 LIST OF TABL ES (Cont'd)

NUMBER TITLE PAGE 5.4-15 Reactor Cool ant System Valve s NDE During Fabrication 5.4-90 5.4-16 Pressurizer Valves Design Parameters 5.4-91 5.4-17 Failure Mode and Eff ects Analysis - Residual Heat Removal System Ac tive Components - Plant Cooldown Operation 5.4-92 5.4-18 Single Failure Evaluation of S ystems Required to Reach Cold Shutdown per PTP RSB 5-1 5.4-101 5.4-19 Summary of Systems and E quipment Required for Cold Shutdown Boration Without Letdown 5.4-105 5.4-20 Deleted 5.4-21 Deleted 5.4-22 Comparison of Hydraulic Resistance Coefficients 5.4-108 5.4-23 Comparison of Up per Head Region Hydraulic Resistance 5.4-109

B/B-UFSAR 5.0-viii REVISION 9 - DECEMBER 2002 CHAPTER 5.0 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS LIST OF FIGURES NUMBER TITLE 5.1-1 Deleted 5.1-2 Reactor Coolant Syst em Process Flow Diagram 5.1-3 Deleted 5.1-4 Deleted 5.2-1 Deleted 5.2-2 Deleted 5.2-3 Deleted 5.2-4 Deleted 5.2-5 Deleted 5.3-1 Reactor Vessel 5.4-1 Reactor Coolant Cont rolled Leakage Pump 5.4-2 Reactor Coolant Pump Estimated Performance Characteristic 5.4-3 K ID Lower Bound F racture Toughness A533V (Reference WCAP 7623) Grade B Class 1

5.4-4 Deleted 5.4-5 Pressurizer 5.4-6 Reactor Coolant Temper ature vs. Time (Normal Cooldown) 5.4-7 Single RHR Train RC Temperature vs. Time 5.4-8 Pressurizer Relief Tank 5.4-9 Reactor Coolant Loop Stop Valve 5.4-10 Unit 1 S team Generator 5.4-11 Unit 2 S team Generator

B/B-UFSAR 5.0-ix REVISION 9 - DECEMBER 2002 CHAPTER 5.0 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS DRAWINGS CITED IN THIS CHAPTER*

  • The listed drawings are included as "General References" only; i.e., refer to the drawings to obtain additional d etail or to obtain background information.

These drawings are not part of the UFSAR.

They are controlled by the C ontrolled Documents Program.

DRAWING* SUBJECT 108D685-11 Pressurizer Pressure a nd Level Contr ol Diagram A-333 Containment Building Bas ement Floor Plan Area 1 A-701 Containment Building Plumbing Diagram M-35 Diagram of Main Stea m System Unit 1 M-60 Diagram of Reactor Coola nt System Loops Unit 1 M-61 Diagram of Safety Inje ction System Unit 1 M-62 Diagram of Residual Heat Removal System Unit 1 M-64 Diagram of Chemical and Volume Control S ystem and Boron Thermal Regeneration System Unit 1 M-64A Diagram of Chemical and Volume Control S ystem and Boron Thermal Regeneration System Unit 1 M-135 Diagram of Reactor Coola nt System Loops Unit 2 M-196 Reactor Coolant System L oop Piping Arrangement S-1066 Containment Building S ections and Details

B/B-UFSAR 5.1-1 REVISION 9 - DECEMBER 2002 CHAPTER 5.0 - REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1

SUMMARY

DESCRIPTION The reactor coolant system (RCS) shown in Draw ings M-60 and M-135 and Figure 5.1-2 consists of s imilar heat transfer loops connected in parallel to the rea ctor pressure ve ssel. Each loop contains a reactor coolant pum p, steam generator and associated piping and valves.

In addition, the system includes a pressurizer, a p ressurizer relief tank, interconnecting piping and instrumentation nece ssary for operationa l control; all of these components are located in the containment building.

During operation, the RCS transfers the heat generated in the core to the steam gene rators where steam is produced to drive the turbine generator. Bora ted water is circulated in the RCS at a flow rate and temperature consis tent with achi eving the reactor core thermal-hydraulic performance. The water also acts as a neutron moderator and re flector, and as a solv ent for the neutron absorber used in chemi cal shim control.

The RCS pressure boundary provides a barrier against the release of radioactivity generated within the reactor and is designed to ensure a high degree of integrity throughout the life of the plant. RCS pressure is contro lled by the use of the pressurizer where water and steam are maintained in equi librium by electrical heaters and water sprays. Steam can be formed (by the heaters) or condensed (by the pressuriz er spray) to min imize pressure variations due to co ntraction and expans ion of the reactor coolant. Spring-loaded safety valves and po wer-operated relief valves are mounted on the pr essurizer and di scharge to the pressurizer relief tan k, where the steam is condensed and cooled by mixing with water.

The extent of the RCS is defined as:

a. the reactor vessel inc luding control rod drive mechanism housings;
b. the reactor coolant side of the steam generators;
c. reactor coolant pumps;
d. a pressurizer at tached to one of the reactor coolant loops; e. the pressuri zer relief tank;
f. safety and relief valves;

B/B-UFSAR 5.1-2 REVISION 1 - DECEMBER 1989 g. the interconnecting pipi ng, valves a nd fittings between the principa l components lis ted previously; and h. the piping, fittings, and valves leading to connecting auxiliary or support systems up to and including the second isolati on valve (fr om the high pressure side) on each line.

Reactor Coolant System Components Reactor Vessel The reactor vessel is cylindrica l, with a weld ed hemispherical bottom head and a remova ble, flanged and gas keted, hemisperhical upper head. The vessel contains the cor e, core supporting structures, control rods, and ot her parts direct ly associated with the core.

The vessel has inlet and outlet nozzles loca ted in a horizontal plane just below the reactor ves sel flange but above the top of the core. Coolant enters the vessel through the inlet nozzles and flows down the core barrel-vessel wall ann ulus, turns at the bottom and flows up th rough the core to the outlet nozzles.

Steam Generators The steam generators are vertical shell and U-tu be evaporators with integral moisture s eparating equipment. The reactor coolant flows through the inverted U-tub es, entering and leaving through the nozzles located in t he hemispherical bottom head of the steam generator. Steam is g enerated on the shell si de and flows upward through the moisture s eparators to t he outlet nozzle at the top of the steam generator.

Reactor Coolant Pumps The reactor coolant pumps are id entical single-speed centrifugal units driven by air-cooled, three-phase indu ction motors. The shaft is vertical with the motor mounted above the pumps. A flywheel on the shaft ab ove the motor provides a dditional inertia to extend pump coastdo wn. The inlet is at the bottom of the pump; discharge is on the side.

Piping The reactor coolant loop piping is specified in sizes consistent with system requirements.

The hot leg inside diameter is 29 inches and the cold leg return line to the reactor vess el is 27-1/2 inches. The piping between the steam generator and the pu mp suction is increased to 31 inches in diameter to reduce pressure dr op and improve flow conditions to the pump suction.

B/B-UFSAR 5.1-3 REVISION 7 - DECEMBER 1998 Pressurizer The pressurizer is a vertical, cylindric al vessel with hemispherical top and bottom h eads. Electri cal heaters are installed through the bottom head of the vessel while the spray nozzle, relief valve, and safety valve connect ions are located in the top head of the vessel.

Pressurizer Relief Tank

The pressurizer relief t ank is a horizontal, cylindrical vessel with elliptical dished h eads. Steam from th e pressurizer safety and relief valves is discharged into the pre ssurizer relief tank through a sparger pipe under the water level.

This condenses and cools the steam by mix ing it with water that is near ambient temperature. To prevent exceedi ng its design pr essure, the tank is equipped with two rupture discs siz ed to accommodate the combined capacity of the safety valves.

Safety and Rel ief Valves The pressurizer safety valves are of the totally enclosed pop-type. The valves are sp ring-loaded, self-activated with backpressure compensat ion. The power-op erated relief valves limit system pressure for large power mi smatch. They are operated automatically or by rem ote manual contr ol. Remotely operated valves are pr ovided to isolate the inlet to the power-operated relief va lves if excessive leakage occurs.

Loop Stop Valves Reactor coolant loop stop valves are r emotely controlled motor-operated gate valves which permit any loop to be isolated from the reactor vessel. The valves on the hot l eg and the cold leg are identical except for the int ernal diameter of the valve ends.

Reactor Coolant System (RCS) Performan ce Characteristics Tabulations of importa nt design and performa nce characteristics of the RCS are provided in Table 5.1

-1. Safety limits and limiting safety system settings are discusse d as part of the Technical Specifications.

Reactor Coolant Flow The reactor coolant flow, a major pa rameter in the design of the system and its c omponents, is established with a detailed design procedure supported by operating plant performance data, by pump model tests and analysis, and by pressure drop tests and analyses of the reactor v essel and fuel assembl ies. Data from all operating plants have indicated that the actual flow has been well above the flow specified for the th ermal design of the B/B-UFSAR 5.1-4 REVISION 9 - DECEMBER 2002 plant. By applying the design procedure described in the following, it is possible to spe cify the expected operating flow with reasonable accuracy.

Three reactor coolant fl ow rates are identif ied for the various plant design considerati ons. The definition s of these flows are presented in the fol lowing paragraphs.

Best Estimate Flow

The best estimate flow is the mo st likely value for the actual plant operating condit ion. This flow is based on the best estimate of the reactor vessel, steam generator and piping flow resistance, and on the best estimate of the reactor coolant pump head-flow capacity, with no uncertai nties assigned to either the system flow resistance or the pump head. System pressure drops, based on best estimate f low, are presented in Table 5.1-1.

Although the best estimate flow is the most li kely value to be expected in operation, more cons ervative flow ra tes are applied in the thermal and mechanical designs.

Thermal Design Flow Thermal design flow is t he basis for the reactor core thermal performance, the steam generator thermal per formance, and the nominal plant parameters used throughout the design. To provide the required margin, the thermal design flow acc ounts for the uncertainties in reactor vessel, ste am generator and piping flow resistances, reactor c oolant pump head, and the methods used to measure flow rate. The thermal design flow is approximately 9.8%

less than best estim ate flow for Byron/B raidwood Units 1 at 5%

steam generator tube plug (SGTP). B yron Unit 2 thermal design flow is 7.5% less th an best estimate flow at 10% SGTP and Braidwood Unit 2 is 6.8% less at 10% SGTP. The thermal design flow is confirmed wh en the plant is plac ed in operation.

Tabulations of importa nt design and performa nce characteristics of the reactor c oolant systems, as provi ded in Table 5.1-1, are based on the thermal design flow.

The minimum acceptable m argin between thermal design loop flow rate and best estimate loop flow rate is 4% for Byron/Braidwood.

As indicated above, the actual thermal design flow rate is more than 4% lower than t he best estimate flow rate. Refer to Subsection 4.4.2.9.6 for a dis cussion of the uncertainties of flow rate.

Mechanical Design Flow Mechanical design flow is the conservatively high flow used in the mechanical design of the reactor vessel internals and fuel assemblies. To ensure that a conservatively high flow is specified, the m echanical design flow is based on a reduced system resistance and on increased pump head capability. The mechanical design flow is approximately 3.9%

for Unit 1 and 5.2%

for Unit 2 greater t han the best estimate flow. Maximum pump overspeed results in a peak reactor cool ant flow of 120% of the mechanical design fl ow. This overspeed condition, which is coincident with a turbine-generator B/B-UFSAR 5.1-5 REVISION 11 - DECEMBER 2006 overspeed of 20%, is only ap plicable if, whe n a turbine trip would be actuated, the turbine governor fails and the turbine is tripped on overspeed.

Interrelated Performance and Safety Functions The interrelated performance and saf ety functions of the RCS and its major components are as follows:

a. The RCS provides suffici ent heat removal capability to transfer the heat produced during and after power operation and when the reactor is subcritical, including the initia l phase of plant c ooldown, to the steam and power conv ersion system.
b. The system provides sufficient heat removal capability to transf er the heat during the subsequent phase of plant coold own and cold shutdown to the residual heat removal system.
c. The system h eat removal capabi lity under power operation and normal o perational transients, including the transition from forced to natural circulation, assures no fuel damage within the operating bounds permitted by the reactor control and protection systems.
d. The RCS contains the wat er used as t he core neutron moderator and reflec tor and as a solve nt for chemical shim control.
e. The system mai ntains the homogen eity of soluble neutron poison conce ntration and rate of change of coolant temperature such that uncontrolled reactivity changes do n ot occur.
f. The reactor vessel is an integral part of the RCS pressure boundary and is cap able of accommodating the temperatures and pressur es associated with the operational transients.

The reactor vessel functions to support the reactor core and control rod drive mechanisms.

g. The pressurizer maintains the system p ressure during operation and limits pressure transients. During the reduction or increase of pla nt load, rea ctor coolant volume changes are accommodated in the pressurizer via the surge line c onnected to the hot leg of one of the reactor coolant loops.

Pressurizer spray is provided via connections to the cold legs of two separate loops.

h. The reactor coolant pumps supply the coolant flow necessary to remove heat fro m the reactor core and transfer it to the steam generators.

B/B-UFSAR 5.1-6 REVISION 9 - DECEMBER 2002 i. The steam genera tors provide steam to the turbine.

The tube and tube sheet boundary are designed to prevent the transfer of activity generat ed within the core to the se condary system.

j. The RCS piping s erves as a bound ary for containing the coolant under operating temperature and pressure conditions and for limit ing leakage (and activity release) to the containment atmosphere. The RCS piping contains bora ted water which is circulated at the flow rate and temp erature consistent with achieving the reactor core thermal and hydraulic performance.

5.1.1 Schematic

Flow Diagram The reactor coolant system is sh own schematically in Figure 5.1-2. Included with th is figure are tabula tions of principal pressures, temperatures, and the flow rate of the system under normal steady-state fu ll power operating con ditions. These parameters are b ased on the best estimate flow at the pump discharge. RCS volume u nder the above conditions is presented in Table 5.1-1.

5.1.2 Piping

and Instr umentation Diagrams A piping and instrumentation diagram of the reactor coolant system is shown on Fig ure 5.1-1. The diagra m shows the extent of the systems located within the c ontainment, and the points of separation between t he reactor coolant syste m and the secondary (heat utilization) system.

5.1.3 Elevation

Drawings Drawing M-196, Sheets 1 and 2, are elevation drawings showing principal dimensions of the reactor coolant sy stem in relation to the supporting or surrou nding concrete structures.

B/B-UFSAR 5.3-1 5.3 REACTOR VESSEL 5.3.1 Reactor Vessel Materials This section is for purposes of reactor pressure vessel fabrication. 5.3.1.1 Material Specifications Material specifications are in accordance with the ASME Code requirements and are given in Subsection 5.2.3. 5.3.1.2 Special Processes Used for Manufacturing and Fabrication a. The vessel is Seismic Category I and Quality Group A. Design and fabrication of the reactor vessel is carried out in strict accordance with ASME Code,Section III, Class 1 requirements. The head flanges and nozzles are manufactured as forgings. The cylindrical portion of the vessel is made up of several forged shells. The hemispherical heads are made from dished plates. The reactor vessel parts are joined by welding, using the single or multiple wire submerged arc. b. The use of severely sensitized steel as a pressure boundary material has been prohibited and has been eliminated by either a select choice of material or by programming the method of assembly. c. The control rod drive mechanism head adaptor threads and surfaces of the guide studs are chrome plated to prevent possible galling of the mated parts. d. At all locations in the reactor vessel where stainless steel and Inconel are joined, the final joining beads are Inconel weld metal in order to prevent cracking. e. The location of full penetration weld seams in the upper closure head and vessel bottom head are restricted to areas that permit accessibility during inservice inspection. f. The stainless steel clad surfaces are sampled to ensure that composition and delta ferrite requirements are met. g. The procedure qualification for cladding low alloy steel (SA508 Class 2) requires a special evaluation to ensure freedom from underclad cracking.

B/B-UFSAR 5.3-2 5.3.1.3 Special Methods for Nondestructive Examination The examination requirements detailed in the following are in addition to the examination requirements of Section III of the ASME Code. The reactor vessel nondestructive examination (NDE) program is given in Table 5.3-1. 5.3.1.3.1 Ultrasonic Examination a. In addition to the design code straight beam ultrasonic test, angle beam inspection of 100% of plate material is performed during fabrication to detect discontinuities that may be undetected by longitudinal wave examination. b. In addition to ASME Section III nondestructive examination, all full penetration welds and heat affected zones in the reactor vessel are ultrasonically examined during fabrication. This test is performed upon completion of the welding and intermediate heat treatment but prior to the final postweld heat treatment. c. The reactor vessel is examined after hydrostatic testing for information. 5.3.1.3.2 Penetrant Examinations The partial penetration welds for the control rod drive mechanism head adaptors and the bottom instrumentation tubes are inspected by dye penetrant after the root pass in addition to code requirements. Core support block attachment welds were inspected by dye penetrant after first layer of weld metal and after each 1/2 inch of weld metal. All clad surfaces and other vessel and head internal surfaces were inspected by dye penetrant after the hydrostatic test. 5.3.1.3.3 Magnetic Particle Examination All magnetic particle examinations of materials and welds were performed in accordance with the following: a. Prior to the final postweld heat treatment - by the prod, coil, or direct contact method. b. After the final postweld heat treatment - by the yoke method. The following surfaces and welds were examined by magnetic particle methods.

B/B-UFSAR 5.3-3 REVISION 8 - DECEMBER 2000 Surface Examinations a. Magnetic particle examination of all exterior vessel and head surfaces after the hydrostatic test. b. Magnetic particle examination of all exterior closure stud surfaces and all nut surfaces after final machining or rolling. Continuous circular and longitudinal magnetization were used. c. Magnetic particle examination of all inside diameter surfaces of carbon and low alloy steel products that have their properties enhanced by accelerated cooling. This inspection is performed after forming and machining (if required) and prior to cladding. Weld Examination Magnetic particle examination of the weld metal buildup for vessel welds attaching the closure head lifting lugs to the reactor vessel after the first layer and each 1/2 inch of weld metal is deposited. All pressure boundary welds were examined after back chipping or back grinding operations. 5.3.1.4 Special Controls for Ferritic and Austenitic Stainless Steels Welding of ferritic steels and austenitic stainless steels is discussed in Subsection 5.2.3. Subsection 5.2.3 includes discussions which indicate the degree of acceptance with guidelines for control of ferrite content in stainless steel metal welds, use of sensitized stainless steel, electroslag weld properties, stainless steel weld cladding of low-alloy steel components and welder qualification for areas of limited accessibility. Appendix A discusses the degree of conformance with regulatory guides. 5.3.1.5 Fracture Toughness Assurance of adequate fracture toughness of ferritic materials in the reactor coolant pressure boundary (ASME Section III Class 1 Components) is provided by compliance with the requirements for fracture toughness testing included in NB-2300 to Section III of the ASME Boiler and Pressure Vessel Code, and Appendix G of 10 CFR 50. The initial Charpy V-notch minimum upper shelf fracture energy levels for the reactor vessel beltline (including welds) shall be 75 foot-pounds as required by Appendix H of 10 CFR 50. Materials having a section thickness greater than 10 inches with an upper shelf of less than 75 foot-pounds shall be evaluated with regard to effects of chemistry (especially copper content), initial upper shelf energy and influence to B/B-UFSAR 5.3-4 REVISION 12 - DECEMBER 2008 ensure that a 50 foot-pound shelf energy as required by Appendix H of 10 CFR 50 is maintained throughout the life of the vessel. The specimens shall be oriented as required by NB-2300 of Section III of the ASME Boiler and Pressure Vessel Code. The reactor vessel material properties for units of the Byron/Braidwood Stations are given in Section 5 of the PTLR and the Bases for Technical Specification 3.4.3. 5.3.1.5.1 Pressurized Thermal Shock Evaluation Fracture toughness requirements for protection of reactor vessels against pressurized thermal shock events are given in 10 CFR 50.61. An evaluation has been performed in accordance with these requirements for the reactor vessels at Byron/ Braidwood Units 1 and 2. The evaluation is provided in References 1 to 4 and the evaluation results are summarized in Tables 5.3-7 through 5.3-10. 5.3.1.6 Material Surveillance In the surveillance program, the evaluation of the radiation damage is based on preirradiation testing of Charpy V-notch and tensile specimens and postirradiation testing of Charpy V-notch, tensile and 1/2 thickness (T) compact tension (CT) fracture mechanics test specimens. The program is directed toward evaluation of the effect of radiation on the fracture toughness of reactor vessel steels based on the transition temperature approach and the fracture mechanics approach. The program conforms with ASTM-E-185 "Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," and 10 CFR 50, Appendix H. Detailed information on the reactor vessel material surveillance program is provided in Westinghouse reports WCAP-9517 for Byron Unit 1, WCAP-10398 for Byron Unit 2, and WCAP-9807 for Braidwood Unit 1 and WCAP-11188 for Braidwood 2. The reactor vessel surveillance program uses six specimen capsules. The capsules are located in guide baskets welded to the outside of the neutron shield pads and are positioned directly opposite the center portion of the core. The capsules can be removed when the vessel head is removed and can be replaced when the internals are removed. The six capsules contain reactor vessel steel specimens, oriented both parallel and normal (longitudinal and transverse) to the principal working direction of the limiting base material located in the core region of the reactor vessel and associated weld metal and weld heat-affected zone metal. The 6 capsules contain 54 tensile specimens, 360 Charpy V-notch specimens (which include weld metal and weld heat-affected zone material), and 72 CT specimens. Archive material sufficient for two additional capsules will be retained.

B/B-UFSAR 5.3-5 REVISION 1 - DECEMBER 1989 Dosimeters, including Ni, Cu, Fe, Co-Al, Cd shielded Co-Al, Cd shielded Np-237 and Cd shielded U-238, are placed in filler blocks drilled to contain them. The dosimeters permit evaluation of the flux seen by the specimens and the vessel wall. In addition, thermal monitors made of low melting point alloys are included to monitor the maximum temperature of the specimens. The specimens are enclosed in a tight-fitting stainless steel sheath to prevent corrosion and ensure good thermal conductivity. The complete capsule is helium leak tested. Each of the six capsules contains the following specimens: Number of Number of Number of Material Charpys Tensiles CTs Limiting base material* 15 3 4 Limiting base material** 15 3 4 Weld metal*** 15 3 4 Heat affected zone 15

  • Specimens oriented in the major working direction. ** Specimens oriented normal to the major working direction. *** Weld metal to be selected per ASTM E185. The following dosimeters and thermal monitors are included in each of the six capsules: Dosimeters Iron Copper Nickel Cobalt-Aluminum (0.15% Co) Cobalt-Aluminum (Cadmium shielded) U-238 (Cadmium shielded) Np-237 (Cadmium shielded) Thermal Monitors 97.5% Pb, 2.5% Ag (579F melting point). 97.5% Pb, 1.75% Ag, 0.75% Sn (590F melting point).

B/B-UFSAR 5.3-6 REVISION 12 - DECEMBER 2008 The fast neutron exposure of the specimens occurs at a faster rate than that experienced by the vessel wall, with the specimens being located between the core and the vessel. Since these specimens experience accelerated exposure and are actual samples from the materials used in the vessel, the transition temperature shift measurements are representative of the vessel at a later time in life. Data from CT fracture toughness specimens are expected to provide additional information for use in determining allowable stresses for irradiated material. Correlations between the calculations and the measurements of the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner wall, are described in Subsection 5.3.1.6.1. They have indicated good agreement. The anticipated degree to which the specimens will perturb the fast neutron flux and energy distribution will be considered in the evaluation of the surveillance specimen data. Verification and possible readjustment of the calculated wall exposure will be made by use of data on all capsules withdrawn. For the schedule for removal of the capsules for postirradiation testing which follows that of 10 CFR 50 Appendix H, refer to Table 4.1 of the PTLR. 5.3.1.6.1 Measurement of Integrated Fast Neutron (E>1.0 MeV) Flux at the Irradiation Samples The use of passive neutron sensors such as those included in the internal surveillance capsule dosimetry sets dose not yield a direct measure of the energy dependent neutron flux level at the measurement location. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average flux level and, hence, time integrated exposure (fluence) experienced by the sensors may be developed from the measurements only if the sensor characteristics and the parameters of the irradiation are well known. In particular, the following variables are of interest: 1. The measured specific activity of each sensor 2. The physical characteristics of each sensor 3. The operating history of the reactor 4. The energy response of each sensor 5. The neutron energy spectrum at the sensor location In this section the procedures used to determine sensor specific activities, to develop reaction rates for individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described.

B/B-UFSAR 5.3-7 REVISION 9 - DECEMBER 2002 5.3.1.6.1.1 DETERMINATION OF SENSOR REACTION RATES The specific activity of each of the radiometric sensors is determined using established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor is determined by means of a high purity germanium gamma spectrometer. In the case of the surveillance capsule multiple foil sensor sets, these analyses are performed by direct counting of each of the individual wires; or, as in the case of U-238 and Np-237 fission monitors, by direct counting preceded by dissolution and chemical separation of cesium from the sensor. The irradiation history of the reactor over its operating lifetime is determined from plant power generation records. In particular, operating data are extracted on a monthly basis from reactor startup to the end of the capsule irradiation period. For the sensor sets utilized in the surveillance capsule irradiations, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. Having the measured specific activities, the operating history of the reactor, and the physical characteristics of the sensors, reaction rates referenced to full power operation are determined from the following equation: where: A = measured specific activity (dps/gm) R = reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref (rps/nucleus). No = number of target element atoms per gram of sensor. F = weight fraction of the target isotope in the sensor material. Y = number of product atoms produced per reaction. Pj = average core power level during irradation period j (MW). Pref = maximum or reference core power level of the reactor (MW). djttjrefjjoeeCPPYFNAR1 B/B-UFSAR 5.3-7a REVISION 9 - DECEMBER 2002 Cj = calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period. = decay constant of the product isotope (sec-1). tj = length of irradiation period j (sec). td = decay time following irradiation period j (sec). and the summation is carried out over the total number of monthly intervals comprising the total irradiation period. In the above equation, the ratio Pj/Pref accounts for month by month variation of power level within a given fuel cycle. The ratio Cj is calculated for each fuel cycle and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel cycle. For a single cycle irradiation Cj = 1.0. However, for multiple cycle irradiations, particularly those employing low leakage fuel management the additional Cj correction must be utilized. 5.3.1.6.1.2 Corrections to Reaction Rate Data Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 5.4.3.6.1.3, additional corrections are made to the U-238 measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. In addition to the corrections made for the presence of U-235 in the U-238 fission sensors, corrections are also made to both the U-238 and Np-237 sensor reaction rates to account for gamma ray induced fission reactions occurring over the course of the irradiation. 5.3.1.6.1.3 Least Squares Adjustment Procedure Least squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimate for key exposure parameters such as (E > 1.0 eV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. The use of measurements in combination with the analytical results reduces the uncertainty in the calculated spectrum and acts to remove biases that may be present in the analytical technique.

B/B-UFSAR 5.3-7b REVISION 9 - DECEMBER 2002 In general, the least squares methods, as applied to pressure vessel fluence evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, relates a set of measured reaction rates, Ri, to a single neutron spectrum g, through the multigroup dosimeter cross-section, ig, each with an uncertainty . The use of least squares adjustment methods in LWR dosimetry evaluations is not new. The American Society for Testing and Materials (ASTM) has addressed the use of adjustment codes in ASTM Standard E944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" and many industry workshops have been held to discuss the various applications. For example, the ASTM-EURATOM Symposia on Reactor Dosimetry holds workshops on neutron spectrum unfolding and adjustment techniques at each of its bi-annual conferences. Th primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement. The analytical method alone may be deficient because it inherently contains uncertainty due to the input assumptions to the calculation. Typically these assumptions include parameters such as the temperature of the water in the peripheral fuel assemblies, by-pass region, and downcomer regions, component dimensions, and peripheral core source. Industry consensus indicates that the use of calculation alone results in overall uncertainties in the neutron exposure parameters in the range of 15-20% (1). By combining the calculated results with available measurements, the uncertainties associated with the key neutron exposure parameters can be reduced. Specifically ASTM Standard E 944 states; "The algorithims of the adjustment codes tend to decrease the variances of the adjusted data compared to the corresponding input values. The least squares adjustment codes yield estimates for the output data with minimum variances, that is, the "best estimates". This is the primary reason for using these adjustment procedures". ASTM E 944 provides a comprehensive listing of available adjustment codes. The FERRET least squares adjustment code (Reference 5) was initially developed at the Hanford Engineering Development Laboratory (HEDL) and has had extensive use in both the Liquid Metal Fast Breeder (LMFBR) program and the NRC Sponsored Light Water Reactor Dosimetry Improvement Program (LWR-PV-SDIP). As a result of participation in several cooperative efforts associated with the LWR-PV-SDIP, the FERRET approach was adopted by Westinghouse in gigigiggRiR B/B-UFSAR 5.3-7c REVISION 9 - DECEMBER 2002 the mid 1980's as the preferred approach for the evaluation of LWR surveillance dosimetry. The least squares methodology was judged superior to the previously employed spectrum averaged cross-section approach that is totally dependent on the accuracy of the shape of the calculated neutron spectrum at the measurement locations. The FERRET code is employed to combine the results of plant specific neutron transport calculations and multiple foil reaction rate measurements to determine best estimate values of exposure parameters ( (E > 1.0 MeV) and dpa) along with associated uncertainties at the measurement locations. The application of the least squares methodology requires the following input: 1. The calculated neutron energy spectrum and associated uncertainties at the measurement location. 2. The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set. 3. The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set. For a given application, the calculated neutron spectrum is obtained from the results of plant specific neutron transport calculations applicable to the irradiation period experienced by the dosimetry sensor set. This calculation is performed using the benchmarked transport calculational methodology described in Section 5.4.3.6.2. The sensor reaction rates are derived from the measured specific activities obtained from the counting laboratory using the specific irradiation history of the sensor set to perform the radioactive decay corrections. The dosimetry reaction cross-sections and uncertainties are obtained from the SNLRML dosimetry cross-section library (Reference 6). The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E1018, "Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB)". There are no additional data or data libraries built into the FERRET code system. All of the required input is supplied externally at the time of the analysis. The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum are input to the least squares procedure in the form of variances and covariances. The assignment of the input uncertainties also follows the guidance provided in ASTM Standard E 944.

B/B-UFSAR 5.3-8 REVISION 9 - DECEMBER 2002 5.3.1.6.2 Calculation of Integrated Fast Neutron (E > 1.0 MeV) Flux at the Irradiation Samples Fast neutron exposure calculations for the reactor geometry are carried out using both forward and adjoint discrete ordinates transport techniques. A single forward calculation provides the relative energy distribution of neutrons for use as input to neutron dosimetry evaluations as well as for use in relating measurement results to the actual exposure at key locations in the pressure vessel wall. A series of adjoint calculations, on the other hand, establish the means to compute absolute exposure rate values using fuel cycle specific core power distributions; thus, providing a direct comparison with all dosimetry results obtained over the operating history of the reactor. In combination, the absolute cycle specific data from the adjoint evaluations together with relative neutron energy spectra distributions from the forward calculation provided the means to: 1. Evaluate neutron dosimetry from surveillance capsule locations. 2. Enable a direct comparison of analytical prediction with measurement. 3. Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves. 5.3.1.6.2.1 Reference Forward Calculation The forward transport calculation for the reactor is carried out in r, geometry using the DORT two-dimensional discrete ordinates code (Reference 7) and the BUGLE-96 cross-section library (Reference 8). The BUGLE-96 library is a 47 neutron group, ENDFB-VI based, data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering is treated with a P3 expansion of the scattering cross-sections and the angular discretization is modeled with an S8 order of angular quadrature. The reference forward calculation is normalized to a core midplane power density characteristic of operation at the stretch rating for the reactor. The spatial core power distribution utilized in the reference forward calculation is derived from statistical studies of long-term operation of Westinghouse 4-loop plants. Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery. Furthermore, for the peripheral fuel assemblies, a 2 uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power is used. Due to the use of this bounding spatial power distribution, the results from the reference forward calculation establish conservative exposure projections for reactors of this design operating at the stretch rating. Since it is unlikely that B/B-UFSAR 5.3-8a REVISION 9 - DECEMBER 2002 actual reactor operation would result in the implementation of a power distribution at the nominal +2 level for a large number of fuel cycles and, further, because of the widespread implementation of low leakage fuel management strategies, the fuel cycle specific calculations for this reactor will result in exposure rates well below these conservative predictions. 5.3.1.6.2.2 Cycle Specific Adjoint Calculations All adjoint analyses are also carried out using an S8 order of angular quadrature and the P3 cross-section approximation from the BUGLE-96 library. Adjoint source locations are chosen at several key azimuths on the pressure vessel inner radius. In addition, adjoint calculations were carried out for sources positioned at the geometric center of all surveillance capsules. Again, these calculations are fun in r, geometry to provide neutron source distribution importance functions for the exposure parameter of interest; in this case, (E > 1.0 MeV). The importance functions generated from these individual adjoint analyses provide the basis for all absolute exposure projections and comparison with measurement. These importance functions, when combined with cycle specific neutron source distributions, yield absolute predictions of neutron exposure at the locations of interest for each of the operating fuel cycles; and, establish the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles. Having the importance functions and appropriate core source distributions, the response of interest can be calculated as: Where: (R0, 0) = Neutron flux (E > 1.0 MeV) at radius R0 and azimuthal angle 0. I(r,,E) = Adjoint importance function at radius r, azimuthal angle , and neutron source energy E. S(r,,E) = Neutron source strength at core location r, and energy E. It is important to note that the cycle specific neutron source distributions, S(r,,E), utilized with the adjoint improtance functions, I(r,,E), permit the use not only of fuel cycle specific spatial variations of fission rates within the reactor core; but, also allow for the inclusion of the effects of the differing neutron yield per fission and the variation in fission spectrum introduced by the build-in of plutonium isotopes as the burnup of individual fuel assemblies increases. dEddrrErSErIREr),,(),,()(0,0 B/B-UFSAR 5.3-8b REVISION 13 - DECEMBER 2010 5.3.1.7 Reactor Vessel Fasteners The reactor vessel closure studs, nuts, and washers are designed, fabricated, and examined in accordance with the requirements of ASME Section III. The closure studs are fabricated of SA-540, Class 3 Grade B23 material. The closure stud material meets the fracture toughness requirements of ASME Section III, and 10 CFR 50 Appendix G. Representative closure head bolting material properties for the Byron and Braidwood Stations are given in Tables 5.3-3a and b. The guidelines for materials and inspections for vessel closure studs are discussed in Appendix A. Inservice nondestructive examinations are performed in accordance with the station ISI program. The studs, nuts, and washers are removed from the refueling cavity and stored at convenient locations on the containment operating deck prior to removal of the reactor closure head and refueling cavity flooding. Therefore, the reactor closure studs are never exposed to the borated refueling cavity water. Additional protection against the possibility of incurring corrosion effects is ensured by the use of a manganese base phosphate surfacing treatment. (For Byron Unit 2, out of service studs may remain installed in the reactor flange when the refueling cavity is flooded.)

B/B-UFSAR 5.3-9 REVISION 13 - DECEMBER 2010 The stud holes in the reactor flange are sealed with special plugs before removing the reactor closure thus preventing leakage of the borated refueling water into the stud holes. (For Byron Unit 2, out of service studs remaining installed in the reactor flange do not have these special plugs installed, therefore, prior to returning the stud to service, the out of service stud is removed and the stud hole inspected per existing procedures.) 5.3.2 Pressure-Temperature Limits 5.3.2.1 Limit Curves Startup and shutdown operating limitations are based on the properties of the core region materials of the reactor pressure vessel. Actual material property test data are used. The methods outlined in Appendix G to Section XI of the ASME Code are employed for the shell regions in the analysis of protection against nonductile failure. The initial operating curves are calculated assuming a period of reactor operation such that the beltline material will be limiting. The heatup and cooldown curves are given in Figures 2.1, 2.2 and Table 2.1 of each station's Pressure Temperature Limits Report (PTLR). Beltline material properties change with radiation exposure, and this change is measured in terms of the adjusted reference nil ductility temperature which includes a reference nil ductility temperature shift (RTNDT). Predicted RTNDT values are derived based on predicted neutron fluence at the assumed vessel wall flaw locations and the methodology provided in Regulatory Guide 1.99, Revision 2. The expected neutron fluence for reactor vessel wall locations of 1/4 T (thickness) and 3/4 T are determined. These reactor vessel wall locations represent the tips of the code reference flaw when the flaw is assumed at the inside diameter and outside diameter locations, respectively. The methodology provided within Regulatory Guide 1.99, Revision 2 is used to calculate RTNDT based on the effects of neutron fluence and the effects of chemical composition of the vessel wall material (specifically, copper and nickel). For a selected time of operation, this shift is assigned a sufficient magnitude so that no unirradiated ferritic materials in other components of the reactor coolant system will be limiting in the analysis. The operating curves including pressure-temperature limitations, are calculated in accordance with 10 CFR 50, Appendix G, and ASME Code Section XI, Appendix G requirements. In addition, Byron Units 1 and 2 and Braidwood Units 1 and 2 have received exemptions from the NRC which permit use of the safety margins recommended in the 1996 Addenda to ASME,Section XI, Appendix G requirements. The optional use of the 1996 Addenda allows for a 10% relaxation in the allowable pressure-temperature curves in the PTLR. Changes in fracture toughness of the core region plates or forgings, weldments and associated heat affected zones due to radiation damage will be monitored by a surveillance program which conforms with ASTM E-185, "Recommended Practice for Surveillance B/B-UFSAR 5.3-9a REVISION 7 - DECEMBER 1998 Tests for Nuclear Reactor Vessels," and 10 CFR 50, Appendix H. Byron and Braidwood Stations have received permission from the NRC to integrate the reactor vessel surveillance programs per 10CFR50, Appendix H, Section III.C. This allows the surveillance programs to be integrated for Byron Units 1 and 2, and Braidwood Units 1 and 2, respectively. The evaluation of the radiation damage in this surveillance program is based on preirradiation testing of Charpy V-notch and tensile specimens and postirradiation testing of Charpy V-notch, tensile, and 1/2 T compact tension specimens. The postirradiation testing will be carried out during the lifetime of the reactor vessel. Specimens are irradiated in capsules B/B-UFSAR 5.3-10 REVISION 12 - DECEMBER 2008 located near the core midheight and removable from the vessel at specified intervals. The results of the radiation surveillance program will be used to verify that the RTNDT predicted from the effects of the fluence, or copper and nickel content is appropriate and to make any changes necessary to correct the fluence, or copper and nickel content if RTNDT determined from the surveillance program is greater or less than the predicted RTNDT. Temperature limits for preservice hydrotests and inservice leak and hydrotests were calculated in accordance with 10 CFR 50, Appendix G. The surveillance program withdrawal schedule is contained in Table 4.1 of the PTLR document for each unit, respectively. Changes to the withdrawal schedule may be made as part of an update to the PTLR under the provisions of 10CFR50.59. Regulatory guides are discussed in Appendix A. 5.3.2.2 Operating Procedures The transient conditions that are considered in the design of the reactor vessel are presented in Subsection 3.9.1.1. These transients are representative of the operating conditions that should prudently be considered to occur during plant operation. The transients selected form a conservative basis for evaluation of the RCS to ensure the integrity of the RCS equipment. Those transients listed as upset condition transients are listed in Table 3.9-1. None of these transients will result in pressure-temperature changes which exceed the heatup and cooldown limitations as described in Subsection 5.3.2.1 and in the Pressure Temperature Limits Report (PTLR). 5.3.3 Reactor Vessel Integrity 5.3.3.1 Design The reactor vessel is cylindrical with a welded hemispherical bottom head and removable, bolted, flanged, and gasketed, hemispherical upper head. The reactor vessel flange and head are sealed by two hollow metallic O-rings. Seal leakage is detected by means of two leakoff paths: one between the inner and outer ring, and one outside the outer O-ring. The vessel contains the core, core support structures, control rods, and other parts directly associated with the core. The reactor vessel closure head contains head adapters. These head adapters are tubular members, attached by partial penetration welds to the underside of the closure head. The upper end of these adapters contain acme threads for the assembly of control rod drive mechanisms or instrumentation adapters. The seal B/B-UFSAR 5.3-10a REVISION 7 - DECEMBER 1998 arrangement at the upper end of these adapters consists of a welded flexible canopy seal. Inlet and outlet nozzles are located symmetrically around the vessel. Outlet nozzles are arranged on the vessel to facilitate optimum layout of the reactor coolant system equipment. The inlet nozzles are tapered from the coolant loop vessel interfaces to the vessel inside wall to reduce loop pressure drop.

B/B-UFSAR 5.3-11 REVISION 8 - DECEMBER 2000 The bottom head of the vessel contains penetration nozzles for connection and entry of the nuclear incore instrumentation. Each nozzle consists of a tubular member made of either an Inconel or an Inconel-stainless steel composite tube. Each tube is attached to the inside of the bottom head by a partial penetration weld. Internal surfaces of the vessel which are in contact with primary coolant are weld overlay with 0.125 inch minimum of stainless steel or Inconel. The exterior of the reactor vessel is insulated with canned stainless steel reflective sheets. The insulation is a minimum of 3 inches thick and contoured to enclose the top, sides, and bottom of the vessel. All the insulation modules are removable but the access to vessel side insulation is limited by the surrounding concrete. The reactor vessel is designed and fabricated in accordance with the requirements of ASME Section III. Principal design parameters of the reactor vessel are given in Table 5.3-2. The vessel is shown in Figure 5.3-1. Cyclic loads are introduced by normal power changes, reactor trip, and startup and shutdown operations. These design base cycles are selected for fatigue evaluation and constitute a conservative design envelope for the projected plant life. Vessel analysis result in a usage factor that is less than 1. The design specifications require analysis to prove that the vessel is in compliance with the fatigue and stress limits of ASME Section III. The loading and transients specified for the analysis are based on the most severe conditions expected during service. The heatup and cooldown rates imposed by plant operating limits are provided in the Pressure and Temperature Limits Report (PTLR). These rates are reflected in the vessel design specifications. 5.3.3.2 Materials of Construction The materials used in the fabrication of the reactor vessel are discussed in Subsection 5.2.3. 5.3.3.3 Fabrication Methods The fabrication methods used in the construction of the reactor vessel are discussed in Subsection 5.3.1.2. 5.3.3.4 Inspection Requirements The inspection methods used in conjunction with the fabrication of the reactor vessel are described in Subsection 5.3.1.3.

B/B-UFSAR 5.3-12 REVISION 8 - DECEMBER 2000 5.3.3.5 Shipment and Installation The reactor vessel was shipped in a horizontal position on a shipping sled with a vessel-lifting truss assembly. All vessel openings were sealed to prevent the entrance of moisture and an adequate quantity of desiccant bags was placed inside the vessel. These were placed in a wire mesh basket attached to the vessel cover. All carbon steel surfaces were painted with a heat resistant paint before shipment except for the vessel support surfaces and the top surface of the external seal ring. The closure head was also shipped with a shipping cover and skid. An enclosure attached to the ventilation shroud support ring protected the control rod mechanism housings. All head openings were sealed to prevent the entrance of moisture and an adequate quantity of desiccant bags were placed inside the head. These were placed in a wire-mesh basket attached to the head cover. All carbon steel surfaces were painted with heat-resistant paint before shipment. A lifting frame was provided for handling the vessel head. 5.3.3.6 Operating Conditions Operating limitations are presented in Subsection 5.3.2 and in the Technical Specifications. The procedures and methods used to ensure the integrity of the reactor vessel under the most severe postulated conditions are described in Subsection 3.9.1.4. 5.3.3.7 Inservice Surveillance The internal surface of the reactor vessel is capable of inspection periodically using visual and/or nondestructive techniques over the accessible areas. During refueling, the vessel cladding is capable of being inspected in certain areas such as the primary coolant outlet nozzles and, if deemed necessary, the core barrel is capable of being removed, making the entire inside vessel surface accessible. The closure head is examined visually in accordance with the requirements of ASME Section XI. Optical devices permit a selective inspection of the cladding, control rod drive mechanism nozzles, and the gasket seating surface. The knuckle transition piece, which is the area of highest stress of the closure head, is accessible on the outer surface for visual inspection, dye penetrant or magnetic particle, and ultrasonic testing. The closure studs can be inspected periodically using visual, magnetic particle and/or ultrasonic techniques.

B/B-UFSAR 5.3-13 The full penetration welds in the following areas of the installed irradiated reactor vessel are available for visual and/or nondestructive inspection: a. Vessel shell - from the inside surface. b. Primary coolant nozzles - from the inside surface. c. Closure head - from the inside and outside surfaces. d. Closure studs, nuts, and washers. e. Field welds between the reactor vessel, nozzles, and the main coolant piping. f. Vessel flange seal surface. The design considerations which have been incorporated into the system design to permit the above inspection are as follows: a. All reactor internals are completely removable. The tools and storage space required to permit these inspections are provided. b. The closure head is stored dry on the reactor operating deck during refueling to facilitate direct visual inspection. c. All reactor vessel studs, nuts, and washers can be removed to dry storage during refueling. d. Removable plugs are provided in the primary shield. The insulation covering the nozzle welds may be removed. The reactor vessel presents access problems because of the radiation levels and remote underwater accessibility to this component. Because of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufacturing procedures in preparation for the periodic nondestructive tests which are required by the ASME inservice inspection code. These are: a. Shop ultrasonic examinations are performed on all internally clad surfaces to an acceptance and repair standard to assure an adequate cladding bond to allow later ultrasonic testing of the base metal from inside surface. The size of cladding bonding defect allowed is 1/4-inch by 3/4-inch in the region bounded by 2T (T = wall thickness) on both sides of each full penetration pressure boundary weld. Unbounded areas exceeding 0.442 in2 (3/4-inch diameter) in all other regions are rejected.

B/B-UFSAR 5.3-14 REVISION 9 - DECEMBER 2002 b. The design of the reactor vessel shell is a clean, uncluttered cylindrical surface to permit future positioning of the test equipment without obstruction. c. The weld deposited cladding surface on both sides of the welds to be inspected is specifically prepared to ensure meaningful ultrasonic examinations. d. During fabrication, all full penetration pressure boundary welds are ultrasonically examined in addition to Code examinations. e. After the shop hydrostatic testing, selected areas of the reactor vessel are ultrasonically tested and mapped to facilitate the inservice inspection program. The vessel design and construction enables inspection in accordance with ASME Section XI. 5.3.4 References 1. WCAP-15365, Rev 0 "Evaluation of Pressurized Thermal Shock for Braidwood Unit 1", September 2000. 2. WCAP-15381, Rev 0 "Evaluation of Pressurized Thermal Shock for Braidwood Unit 2", September 2000. 3. WCAP-15390, Rev 0 "Evaluation of Pressurized Thermal Shock for Bryon Unit 1" September 2000. 4. WCAP-15389, Rev 0 "Evaluation of Pressurized Thermal Shock for Byron Unit 2, September 2000. 5. Schmittroth, E. A., "FERRET Data Analysis Code", HEDL-TME-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979. 6. RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994. 7. RSICC Computer Code Collection CCC-650, "DOORS 3.2, One- Two- and Three- Dimensional Discrete Ordinates Neutron/Photon Transport Code System", August 1996. 8. RSIC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications", March 1996.

BYRON-UFSAR 5.3-15 TABLE 5.3-1 REACTOR VESSEL NONDESTRUCTIVE EXAMINATION DURING FABRICATION RT* UT* PT* MT* Forgings 1. Flanges yes yes 2. Studs, nuts yes yes 3. Head adapters yes yes 4. Head adapter tube yes yes 5. Instrumentation tube yes yes 6. Main nozzles yes yes 7. Nozzle safe ends yes yes Plates yes yes Weldments 1. Main seam yes yes yes 2. CRDM head adapter tube assembly to RPV yes 3. Instrumentation tube connection yes 4. Main nozzle yes yes yes 5. Cladding yes yes 6. Nozzle safe ends (if forging) yes yes yes 7. Nozzle safe ends (if weld deposit) yes yes yes 8. Head adapter forging to head adapter tube yes yes 9. All ferritic welds accessible after hydrotest yes yes BYRON-UFSAR 5.3-16 TABLE 5.3-1 (Cont'd) RT UT PT MT 10. Certain non-ferritic welds accessible after hydrotest yes yes 11. Seal ledge yes 12. Head lift lugs yes 13. Core pad welds yes ____________________ *Key: RT - Radiographic, UT - Ultrasonic, PT - Dye penetrant, and MT - Magnetic particle.

BYRON-UFSAR 5.3-17 REVISION 13 - DECEMBER 2010 TABLE 5.3-2 REACTOR VESSEL DESIGN PARAMETERS Design/operating pressure, psig 2485/2317 Design temperature, F 650 Overall height of vessel and closure head, ft-in (bottom head outside diameter to top of control rod mechanism adapter) 43-10 Thickness of insulation, minimum, in. 3 Number of reactor closure head/studs 54 (Note 1) Diameter of reactor closure head/studs, in. (minimum shank) 6-3/4 Inside diameter of flange, in. 167 Outside diameter of flange, in. 205 Inside diameter at shell, in. 173 Inlet nozzle inside diameter, in. 27-1/2 Outlet nozzle inside diameter, in. 29 Cladding thickness, minimum, in. 1/8 Lower head thickness, minimum, in. 5-3/8 Vessel belt-line thickness, minimum, in. 8-1/2 Closure head thickness, in. 6-1/2 ____________________ Note 1- Operation with 53 studs is acceptable for Byron Unit 2 and Braidwood Unit 2. Analyses show that the structural integrity of the Reactor Vessel and all of the stress intensity and fatigue usage factor limits of the ASME Code,Section III, 1971 Edition with Addenda through Summer 1973, are satisfied with one (1) stud out of service. Analyses have also been performed for Byron Unit 1 and Braidwood Unit 1 to justify operation with 53 studs. However, such operation has only been implemented for Byron Unit 2 and Braidwood Unit 2.

BYRON-UFSAR 5.3-18 TABLE 5.3-3a BYRON UNIT 1 CLOSURE HEAD BOLTING MATERIAL PROPERTIES CLOSURE HEAD STUDS Lat. 0.2% YSUTS Elong.RA Energy at 4F Expansion Heat No. Grade Bar No.KSI KSI % % BHN FT-LBS MILS 521190 A540, B23 2B1 130.3 150.8 18.0 60.8 341 55-55-54.5 33-30-33 521190 A540, B23 2B1 133.3 150.8 17.0 55.7 375 46-48-48.5 26-29-29 521190 A540, B23 1B5 133.4 152.1 17.5 57.7 363 45-48-45 26-29-26 521190 A540, B23 1B5 140.2 158.6 16.5 55.0 375 46-47-46.5 26-27-26 521187 A540, B23 1B5 139.2 156.1 17.0 56.9 363 61-61-62 33-34-36 521187 A540, B23 1B5 150.2 166.6 15.0 55.9 375 52-50-50 29-29-29 521187 A540, B23 2B5 144.8 160.8 15.5 57.8 341 59.5-60-59.5 33-32-32 521187 A540, B23 2B5 153.0 168.8 15.0 54.6 363 47-48.5-48 26-30-28 719731 A540, B23 B5 132.8 150.4 17.5 59.4 331 66.5-55-66.5 44-44-44 719731 A540, B23 B5 139.0 156.6 17.0 58.5 363 52.5-53-54 31-31-33 521195 A540, B23 B5 132.3 156.3 17.0 54.9 352 56-53.5-56 32-33-31 521195 A540, B23 B5 141.4 161.9 16.0 54.8 375 45-46-45.5 25-26-26 719673 A540, B23 B4 131.3 151.4 17.0 59.4 331 68-66.5-67.5 43-44-45 719673 A540, B23 B4 134.8 156.6 16.0 57.4 363 53-53.5-52.5 32-33-33 CLOSURE HEAD NUTS & WASHERS 6071004 A540, B23 3153A 142.5 161.1 17.9 56.0 341 55-54-52 29-28-28 6071004 A540, B23 3153B 137.5 156.5 19.3 47.2 341 49-50-49 25-26-31 BRAIDWOOD-UFSAR 5.3-19 TABLE 5.3-3b BRAIDWOOD UNIT 1 CLOSURE HEAD BOLTING MATERIAL PROPERTIES CLOSURE HEAD STUDS Lat. 0.2% YSUTS Elong.RA Energy at 4F Expansion Heat No. Grade Bar No.KSI KSI % % BHN FT-LBS MILS 6053761 A540, B23 1A 138.0 152.5 17.0 54.7 331 52, 54, 54 28, 28, 30 6053761 A540, B23 1B 140.0 153.5 18.0 56.5 352 52, 53, 54 31, 26, 29 6053761 A540, B23 2A 137.0 153.0 18.0 54.7 331 56, 49, 57 43, 37, 37 6053761 A540, B23 2B 141.5 156.0 17.0 54.1 352 54, 54, 55 42, 37, 39 6053761 A540, B23 3A 142.5 159.0 16.0 52.5 331 46, 48, 48 28, 27, 28 6053761 A540, B23 3B 137.0 152.5 16.0 54.9 363 49, 48, 50 26, 25, 25 6053761 A540, B23 4A 144.0 160.0 16.0 50.0 331 51, 48, 50 36, 32, 32 6053761 A540, B23 4B 152.0 165.0 15.0 49.2 363 46, 47, 47 28, 27, 28 6053761 A540, B23 5A 140.0 156.0 17.0 50.0 321 48, 52, 54 33, 35, 38 6053761 A540, B23 5B 148.0 162.5 17.0 52.5 363 52, 46, 48 33, 31, 26 6053761 A540, B23 6A 145.0 159.5 18.0 55.5 321 49, 50, 50 31, 32, 29 6053761 A540, B23 6B 146.5 160.0 15.0 58.6 363 49, 51, 51 26, 35, 33 6053761 A540, B23 7A 140.0 155.0 18.0 56.5 341 55, 52, 51 32, 30, 25 6053761 A540, B23 7B 149.5 164.0 16.0 52.8 363 48, 48, 46 28, 32, 27 6053761 A540, B23 8A 143.0 157.5 17.0 56.8 341 45, 50, 47 28, 32, 26 6053761 A540, B23 8B 146.0 161.5 16.0 52.8 363 47, 47, 46 28, 27, 26 214444 A540, B23 1A 142.0 157.5 16.0 56.5 321 52, 53, 53 34, 37, 34 214444 A540, B23 1B 146.0 162.5 15.5 53.3 363 45, 49, 47 25, 31, 25 214444 A540, B23 2A 139.5 153.5 17.0 57.3 321 51, 53, 55 35, 34, 33 214444 A540, B23 2B 136.0 153.0 16.0 53.3 363 50, 48, 47 28, 29, 26 CLOSURE HEAD NUTS 43135 A540, B23 1A 146.9 161.2 18.1 55.1 341 50, 55, 58 38, 40, 42 43135 A540, B23 1B 146.8 163.8 19.9 57.4 341 53, 54, 52 40, 40, 38 CLOSURE HEAD WASHERS 43135 A540, B23 1A 157.1 169.4 19.5 55.1 363 50, 51, 50 33, 40, 40 43135 A540, B23 1B 148.0 162.5 19.5 56.3 363 50, 45, 45 28, 25, 36 B/B-UFSAR 5.3-20 TABLES 5.3-4 THROUGH TABLE 5.3-6 HAVE BEEN DELETED BYRON-UFSAR 5.3-21 REVISION 9 - DECEMBER 2002 TABLE 5.3-7 BYRON UNIT 1 PRESSURIZED THERMAL SHOCK EVALUATION INSIDE CHEMICAL CONSTANTS FOR SURFACE PTS CALCULATED MATERIAL DESCRIPTION COMPOSITION, RTPTS CALCULATIONS, FLUENCE SCREENING RTPTS REACTOR VESSEL HEAT w/o (F) (n/cm2) CRITERIA (F) BELTLINE REGION LOCATION NUMBER TYPE COPPER NICKEL INITIAL RTNDT MARGIN 32 EFPY (F) 32 EFPY Lower Nozzle Belt 123J218 SA 508 C1 2 mod. .05 .72 +30 26.7 6.04E18 270 83 Upper Shell 5P-5933 SA 508 C1 2 mod. .04 .74 +40 34 2.02E19 270 110 Lower Shell 5P-5951 SA 508 C1 2 mod. .04 .64 +10 30.9 2.02E19 270 72 Upper Circumferential WF501 ASA/Linde 80 .03 .67 +10 22.1 6.04E18 300 54 Weld Middle Circumferential WF336 ASA/Linde 80 .04 .63 -30 28 1.94E19 300 76 Weld Lower Circumferential WF472 ASA/Linde 80 .23 .57 +10 -- <E17 300 -- Weld BYRON-UFSAR 5.3-22 REVISION 9 - DECEMBER 2002 TABLE 5.3-8 BYRON UNIT 2 PRESSURIZED THERMAL SHOCK EVALUATION INSIDE CHEMICAL CONSTANTS FOR SURFACE PTS CALCULATEDMATERIAL DESCRIPTION COMPOSITION, RTPTS CALCULATIONS, FLUENCE SCREENING RTPTS REACTOR VESSEL HEAT w/o (F) (n/cm2) CRITERIA (F) BELTLINE REGION LOCATION NUMBER TYPE COPPER NICKEL INITIAL RTNDT MARGIN 32 EFPY (F) 32 EFPY Lower Nozzle Belt 4P-6107 SA 508 C1 2 mod. .05 .74 +10 25.4 5.22E18 270 61 Upper Shell 49D329-1-1 SA 508 C1 3 .01 .70 -20 24 2.06E19 270 28 49C297-1-1 Lower Shell 49D330-1-1 SA 508 C1 3 .05 .73 -20 17 2.06E19 270 19 49C298-1-1 Upper Circumferential WF562 ASA/Linde 80 .03 .65 +40 21 5.22E18 300 82 Weld Middle Circumferential WF447 ASA/Linde 80 .053 .62 +10 28 2.03E19 300 116 Weld Lower Circumferential WF614 ASA/Linde 80 .18 .54 +40 -- <E17 300 -- Weld BRAIDWOOD-UFSAR 5.3-23 REVISION 9 - DECEMBER 2002 TABLE 5.3-9 BRAIDWOOD UNIT 1 PRESSURIZED THERMAL SHOCK EVALUATION INSIDE CHEMICAL CONSTANTS FOR SURFACE PTS CALCULATEDMATERIAL DESCRIPTION COMPOSITION, RTPTS CALCULATIONS, FLUENCE SCREENING RTPTS REACTOR VESSEL HEAT w/o (F) (n/cm2) CRITERIA (F) BELTLINE REGION LOCATION NUMBER TYPE COPPER NICKEL INITIAL RTNDT MARGIN 32 EFPY (F) 32 EFPY Lower Nozzle Belt 5P-7016 SA 508 C1 2 mod. .04 .73 +10 22.4 6.08E18 270 55 Upper Shell 49C344-1-1 SA 508 C1 3 .05 .73 -30 34 2.05E19 270 41 49D383-1-1 Lower Shell 49D867-1-1 SA 508 C1 3 .05 .74 -20 17 2.05E19 270 26 49C813-1-1 Upper Circumferential WF645 ASA/Linde 80 .04 .46 -25 39.6 6.08E18 300 54 Weld Middle Circumferential WF562 ASA/Linde 80 .03 .67 +40 28 1.99E19 300 99 Weld Lower Circumferential WF653 ASA/Linde 80 .19 .58 - <E17 300 -- Weld B/B-UFSAR 5.3-24 REVISION 9 - DECEMBER 2002 TABLE 5.3-10 BRAIDWOOD UNIT 2 PRESSURIZED THERMAL SHOCK EVALUATION INSIDE CHEMICAL CONSTANTS FOR SURFACE PTS CALCULATEDMATERIAL DESCRIPTION COMPOSITION, RTPTS CALCULATIONS, FLUENCE SCREENING RTPTS REACTOR VESSEL HEAT w/o (F) (n/cm2) CRITERIA (F) BELTLINE REGION LOCATION NUMBER TYPE COPPER NICKEL INITIAL RTNDT MARGIN 32 EFPY (F) 32 EFPY Lower Nozzle Belt 5P-7056 SA 508 C1 2 mod. .04 .90 +30 21.9 5.67E18 270 74 Upper Shell 49D963-1-1 SA 508 C1 3 .03 .71 -30 23.6 1.96E19 270 17 49C904-1-1 Lower Shell 50D102-1-1 SA 508 C1 3 .06 .75 -30 34 1.96E19 270 19 50C97-1-1 Upper Circumferential WF645 ASA/Linde 80 .033 .50 -30 45.4 5.67E18 300 66 Weld Middle Circumferential WF562 ASA/Linde 80 .03 .65 +40 28 1.89E19 300 98 Weld Lower Circumferential WF696 ASA/Linde 80 .04 .60 - <E17 300 -- Weld

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