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{{#Wiki_filter:B/B-UFSAR 1.0-i REVISION 5 - DECEMBER 1994 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS
{{#Wiki_filter:B/B-UFSAR 1.0-i REVISION 5 - DECEMBER 1994 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS PAGE  
 
PAGE


==1.0 INTRODUCTION==
==1.0 INTRODUCTION==
AND GENERAL DESCRIPTION OF PLANT 1.1-1
AND GENERAL DESCRIPTION OF PLANT 1.1-1  


==1.1 INTRODUCTION==
==1.1 INTRODUCTION==
1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Site and Environment 1.2-1 1.2.2 Nuclear Steam Supply System 1.2-1 1.2.3 Engineered Safety Features 1.2-2 1.2.4 Emergency Core Cooling System 1.2-3 1.2.5 Control and Instrumentation 1.2-3 1.2.6 Electrical System 1.2-4 1.2.7 Turbine and Auxiliaries 1.2-4 1.2.8 Fuel Handling System 1.2-5 1.2.9 Radioactive Waste Management System 1.2-5 1.2.10 Features of Special Interest 1.2-5 1.2.11 Structures 1.2-6 1.3 COMPARISON TABLES 1.3-1 1.3.1 Comparisons with Similar Facility Designs 1.3-1 1.3.2 Comparison of Final and Preliminary Information 1.3-2 1.3.3 References 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Licensee 1.4-1 1.4.2 Architect-Engineer 1.4-1 1.4.3 Reactor Designer 1.4-1 1.4.4 Constructor 1.4-2 1.4.5 Consultants and Service Organization 1.4-2 1.4.5.1 Security Systems - ETA 1.4-2 1.4.5.2 Dames & Moore 1.4-3 1.4.5.3 HARZA Engineering 1.4-3 1.4.5.4 Murray and Trettel, Inc.
1.4-3 1.4.5.5 Shirmer Engineering Corporation 1.4-3 1.4.5.6 Hyla S. Napadensky 1.4-4 1.4.5.7 NALCO Chemical Company 1.4-4 1.4.5.8 Westinghouse Environmental Systems Department (WESD) 1.4-4 1.4.5.9 Illinois Natural History Survey (INHS) 1.4-5 1.4.5.10 NUS Corporation 1.4-5 1.4.5.11 Eberline Instrument Corporation (EIC) 1.4-5


1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Site and Environment 1.2-1 1.2.2 Nuclear Steam Supply System 1.2-1 1.2.3 Engineered Safety Features 1.2-2 1.2.4 Emergency Core Cooling System 1.2-3 1.2.5 Control and Instrumentation 1.2-3 1.2.6 Electrical System 1.2-4 1.2.7 Turbine and Auxiliaries 1.2-4 1.2.8 Fuel Handling System 1.2-5 1.2.9 Radioactive Waste Management System 1.2-5 1.2.10 Features of Special Interest 1.2-5 1.2.11 Structures 1.2-6
B/B-UFSAR 1.0-ii REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS (Cont'd)
PAGE 1.4.5.12 Meteorology Research, Inc. (MRI) 1.4-5 1.4.5.13 Illinois State Museum (ISM) 1.4-6 1.4.5.14 Equitable Environmental Health, Inc. (EEH) 1.4-6 1.4.5.15 Espey, Huston & Associates, Inc. (EH & A) 1.4-6 1.4.5.16 University of Wisconsin-Milwaukee (UWM) 1.4-7 1.4.5.17 Aero-Metric Engineering, Inc. (AME) 1.4-7 1.4.5.18 Iowa Institute of Hydraulic Research 1.4-7 1.4.5.19 Babcock and Wilcox International (B&W) 1.4-8 1.4.5.20 Framatome Technologies, Incorporated (FTI) 1.4-8 1.4.5.21 Stone & Webster Engineers and Constructors, Inc, (S&W) 1.4-8 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Programs Required For Plant Operation 1.5-1 1.5.1.1 Core Stability Evaluation 1.5-1 1.5.2 Other Programs Not Required For Plant Operation 1.5-1 1.5.2.1 Fuel Development Program For Operation at High Power Densities 1.5-2 1.5.2.2 Blowdown Forces Program 1.5-2 1.5.2.3 Blowdown Heat Transfer Testing 1.5-2 1.5.3 References 1.5-4 1.6 MATERIAL INCORPORATED BY REFERENCES 1.6-1 1.7 DRAWINGS 1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Drawings for Independent Structural Review 1.7-1


1.3 COMPARISON TABLES 1.3-1 1.3.1 Comparisons with Similar Facility Designs 1.3-1
B/B-UFSAR 1.0-iii REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF TABLES NUMBER TITLE PAGE 1.3-1 Plants Using Three-Buttress Containment Design 1.3-3 1.4-1 Exelon Generation Company's Nuclear Power Plants in Service or Under Construction 1.4-9 1.4-2 Nuclear Power Plants Completed or Currently Under Design by Sargent & Lundy 1.4-10 1.4-3 Westinghouse Pressurized Water Reactor Nuclear Power Plants 1.4-11 1.5-1 Blowdown Heat Transfer Phase I Test Parameters 1.5-5 1.5-2 Blowdown Heat Transfer Phase II Test Parameters 1.5-6 1.6-1 Topical Reports Incorporated by Reference 1.6-2 1.7-1 Deleted 1.7-2  
 
1.3.2 Comparison of Final and Preliminary Information 1.3-2 1.3.3 References 1.3-2
 
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Licensee 1.4-1 1.4.2 Architect-Engineer 1.4-1 1.4.3 Reactor Designer 1.4-1 1.4.4 Constructor 1.4-2 1.4.5 Consultants and Service Organization 1.4-2 1.4.5.1 Security Systems - ETA 1.4-2 1.4.5.2 Dames & Moore 1.4-3 1.4.5.3 HARZA Engineering 1.4-3 1.4.5.4 Murray and Trettel, Inc. 1.4-3 1.4.5.5 Shirmer Engineering Corporation 1.4-3 1.4.5.6 Hyla S. Napadensky 1.4-4 1.4.5.7 NALCO Chemical Company 1.4-4 1.4.5.8 Westinghouse Environmental Systems Department (WESD) 1.4-4 1.4.5.9 Illinois Natural History Survey (INHS) 1.4-5 1.4.5.10 NUS Corporation 1.4-5 1.4.5.11 Eberline Instrument Corporation (EIC) 1.4-5
 
B/B-UFSAR 1.0-ii  REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS  (Cont'd)
 
PAGE 1.4.5.12 Meteorology Research, Inc. (MRI) 1.4-5 1.4.5.13 Illinois State Museum (ISM) 1.4-6 1.4.5.14 Equitable Environmental Health, Inc. (EEH) 1.4-6 1.4.5.15 Espey, Huston & Associates, Inc. (EH & A) 1.4-6 1.4.5.16 University of Wisconsin-Milwaukee (UWM) 1.4-7 1.4.5.17 Aero-Metric Engineering, Inc. (AME) 1.4-7 1.4.5.18 Iowa Institute of Hydraulic Research 1.4-7 1.4.5.19 Babcock and Wilcox International (B&W) 1.4-8 1.4.5.20 Framatome Technologies, Incorporated (FTI) 1.4-8 1.4.5.21 Stone & Webster Engineers and Constructors, Inc, (S&W) 1.4-8 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Programs Required For Plant Operation 1.5-1 1.5.1.1 Core Stability Evaluation 1.5-1 1.5.2 Other Programs Not Required For Plant Operation 1.5-1 1.5.2.1 Fuel Development Program For Operation at High Power Densities  1.5-2 1.5.2.2 Blowdown Forces Program 1.5-2 1.5.2.3 Blowdown Heat Transfer Testing 1.5-2 1.5.3 References 1.5-4
 
1.6 MATERIAL INCORPORATED BY REFERENCES 1.6-1 
 
===1.7 DRAWINGS===
1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Drawings for Independent Structural Review 1.7-1
 
B/B-UFSAR 1.0-iii REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF TABLES
 
NUMBER TITLE PAGE 1.3-1 Plants Using Three-Buttress Containment Design 1.3-3 1.4-1 Exelon Generation Company's Nuclear Power Plants in Service or Under Construction 1.4-9 1.4-2 Nuclear Power Plants Completed or Currently Under Design by Sargent & Lundy 1.4-10 1.4-3 Westinghouse Pressurized Water Reactor Nuclear Power Plants 1.4-11 1.5-1 Blowdown Heat Transfer Phase I Test Parameters 1.5-5 1.5-2 Blowdown Heat Transfer Phase II Test Parameters 1.5-6 1.6-1 Topical Reports Incorporated by Reference 1.6-2 1.7-1 Deleted 1.7-2  
 
B/B-UFSAR 1.0-iv  REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 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.


B/B-UFSAR 1.0-iv REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 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.
DRAWINGS*
DRAWINGS*
SUBJECT   M-1 General Site Plan Units 1 & 2 M-2 Property Development Units 1 & 2 M-5 General Arrangement Roof Plan Units 1 & 2 M-6 General Arrangement Main Floor At El. 451'-0" Units 1  
SUBJECT M-1 General Site Plan Units 1 & 2 M-2 Property Development Units 1 & 2 M-5 General Arrangement Roof Plan Units 1 & 2 M-6 General Arrangement Main Floor At El. 451-0 Units 1  
& 2 M-7 General Arrangement Mezzanine Floor At El. 426'-0" Units 1 & 2 M-8 General Arrangement Grade Floor At El. 401'-0" Units 1 & 2 M-9 General Arrangement Floor Plan At El. 383'-0" Units 1  
& 2 M-7 General Arrangement Mezzanine Floor At El. 426-0 Units 1 & 2 M-8 General Arrangement Grade Floor At El. 401-0 Units 1 & 2 M-9 General Arrangement Floor Plan At El. 383-0 Units 1  
& 2 M-10 General Arrangement Basement Floor At El. 364'-0" Units 1 & 2 M-11 General Arrangement Floor Plan At El. 346'-0" Units 1  
& 2 M-10 General Arrangement Basement Floor At El. 364-0 Units 1 & 2 M-11 General Arrangement Floor Plan At El. 346-0 Units 1  
& 2 M-12 General Arrangement Radwaste/Service Building Units 1  
& 2 M-12 General Arrangement Radwaste/Service Building Units 1  
& 2 M-13 General Arrangement Fuel Handling Building Units 1 &
& 2 M-13 General Arrangement Fuel Handling Building Units 1 &
2 M-14 General Arrangement Section "A-A" Units 1 & 2 M-15 General Arrangement Section "B-B" Units 1 & 2 M-16 General Arrangement Section "C-C" and "D-D" Units 1 &
2 M-14 General Arrangement Section A-A Units 1 & 2 M-15 General Arrangement Section B-B Units 1 & 2 M-16 General Arrangement Section C-C and D-D Units 1 &
2 M-17 General Arrangement Section "E-E" Units 1 & 2 M-18 General Arrangement Section "F-F" Units 1 & 2 M-19 General Arrangement Lake Screen House Units 1 & 2 (Braidwood)
2 M-17 General Arrangement Section E-E Units 1 & 2 M-18 General Arrangement Section F-F Units 1 & 2 M-19 General Arrangement Lake Screen House Units 1 & 2 (Braidwood)
M-20 General Arrangement River Screen House Units 1 & 2 M-22 General Arrangement Miscellaneous Plans Units 1 & 2 M-34 P&ID Index and Symbols Units 1 & 2  
M-20 General Arrangement River Screen House Units 1 & 2 M-22 General Arrangement Miscellaneous Plans Units 1 & 2 M-34 P&ID Index and Symbols Units 1 & 2  


B/B-UFSAR 1.3-1 REVISION 8 - DECEMBER 2000 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs
B/B-UFSAR 1.1-1 REVISION 15 - DECEMBER 2014 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT


The design is conceptual ly similar to Exelon Generation Company's Zion Station. D ifferences in the design of the two plants have been allowed only (1) when dic tated by the site characteristics, (2) when the change would result in significant safety improvement, simplif ication of constru ction or operation procedures, or cost savi ngs; or (3) as required to comply with appropriate codes and st andards, NRC criteria, regulatory guides, and regulations.  
==1.1 INTRODUCTION==
 
The Nuclear Regulatory Commission approved the transfer of the facility licenses from Commonwealth Edison (ComEd) Company to Exelon Generation Company, LLC (EGC) on August 3, 2000 (Reference 1). References in the Updated Final Safety Analysis Report (UFSAR) to ComEd, CECo, and Commonwealth Edison have been retained, as appropriate, instead of being changed to EGC to properly preserve the historical context.
The nuclear steam supply system is similar to that of the Zion Station but has a slight ly higher power rati ng. The reactor containments are of the same materials and s ize as those at the Zion Station, but each has only three buttresses, rather than six as used at Zion.
This UFSAR is submitted by Exelon Generation Company for nuclear power plants at Byron, Illinois and at Braidwood, Illinois (Drawings M-1 and M-2) in accordance with the requirements of 10 CFR 50.71(e). Each power plant consists of two units having nearly identical nuclear steam supply systems (NSSS) and turbine generators. The main exception is that the original Unit 1 steam generators were replaced by steam generators of a different design. The power plants at the two sites are as nearly identical as site characteristics permit. The bulk of this UFSAR applies to the standardized, non-site-related aspects of the power plants. Sections which describe features specific to the sites are repeated for each site and the applicable station name appears at the top of these pages. Every effort has been made in the preparation of this document to conform to the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", Revision 2, September 1975. The guidance provided in Nuclear Energy Institute (NEI) 98-03, Guidelines for Updating Final Safety Analysis Reports, Revision 1, June 1999, as endorsed by NRC Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10CFR50.71(e), Revision 0, September 1999, is used to comply with the provisions of 10CFR50.71(e).
The number of post-t ensioning tendons is reduced, and the number of wires per tendon incr eased, from that used at Zion. The r educed number of buttresses allows for greater separation of penetration ar eas for redundant safety-related systems.
Each nuclear power plant consists of two nearly identical generating units, and two pressurized water reactor (PWR) (NSSS) and turbine-generator furnished by Westinghouse Electric Corporation (Westinghouse) similar in design concept to several projects recently licensed or currently under review by the NRC (see Section 1.3). Unit 1 contains steam generators supplied by B&W and Unit 2 contains steam generators supplied by Westinghouse.
Several plants on which this buttress design has been used are listed in Table 1.3-1.
Westinghouse Electric Corporation, Sargent &
The polar cranes in the reactor containment are mounted on the containment wall, rather than on the missile barrier as at Zion.
Lundy, and the Commonwealth Edison Company jointly participated in the original design and construction of each unit. The plant is operated by Exelon Generation Company. Sargent & Lundy (S&L) is the architect-engineer for both stations.
This allows use of a g reater area for co mponent laydown in the containment.  
Each nuclear steam supply system (NSSS) has been evaluated at a power output of 3672 MWt for the Measurement Uncertainty Recapture (MUR) Power Uprate. The warranted gross and approximate net electrical outputs for the MUR are 1268 MWe and 1241 MWe for Unit 1 and Unit 2, respectively. Safety analyses are evaluated at an NSSS power level of 3672 MWt and a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level of 3648 MWt.


Two 100%-capacity cont ainment spray systems are used, rather than the three systems used at Zion.
B/B-UFSAR 1.1-1a REVISION 15 - DECEMBER 2014 Specifically, the containment and engineered safety features (ESF) are designed and evaluated for operation at a core thermal power level of 3658 MWt. Accidents (such as loss-of-coolant, steamline break, and other postulated accidents having offsite dose consequences) are also analyzed at a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level 3648 MWt.
Four containment fan coolers are used, rather than the five u sed at Zion.
The emergency diesel-generator systems for each unit are entir ely independent and use two 5500-kW diesel generators per unit.
The arrangement of equipment in the co mmon auxiliary bui lding allows greater physical separation of r edundant systems and their piping and


cables than was possible at Zion.
B/B-UFSAR 1.1-2 REVISION 9 - DECEMBER 2002 The reactor containments are of post-tensioned concrete construction with a carbon steel liner. Sufficient free volume is provided to contain the energy released in a major accident without need for "pressure suppression" devices. Sargent & Lundy is responsible for containment design.
The Byron Station uses natural draft cooling towers for heat rejection. Zion utilizes once-t hrough cooling.
Byron Station is located in north central Illinois, near the town of Byron and near the Rock River (Drawing M-1). Cooling for the plant is provided by two natural draft cooling towers for non-essential service cooling, and by mechanical draft cooling towers for essential cooling. The fuel loading dates for the two units were November 1984 and November 1986 for Units 1 and 2, respectively. The corresponding dates for commercia1 operation were September 1985 and August 1987.
Mechanical draft cooling towers are provided for essential service cooling at Byron.  
The Braidwood Station is located in northeastern Illinois, near the town of Braidwood and near the Kankakee River (Drawing M-1).
Cooling for the plant is provided by a large man-made cooling pond of approximately 2500 acres constructed over a previously strip-mined area. Essential service cooling is provided by a 99-acre auxiliary cooling pond which is integral with the main pond. The fuel loading dates for the two units were October 1986 and December 1987 for Units 1 and 2, respectively. The corresponding dates for commercial operation were July 1988 and October 1988.
The standard symbols used on piping and instrument diagrams and other figures in this UFSAR are shown in Drawing M-34.


The Braidwood Statio n uses a large man-made cooling pond for heat rejection. An auxiliary cooling pond, integral with the main pond, is provided for es sential service cooling.
==1.2 REFERENCES==
1.
NRC letter, "Braidwood, Byron, Dresden, LaSalle, Quad Cities, and Zion - Orders Approving Transfer of Licenses From Commonwealth Edison Company To Exelon Generation Company, LLC, and Approving Conforming Amendments," dated August 3, 2000


Table 1.3-2 of the FSAR provided the des ign comparison of the Byron/Braidwood nuclear steam supply system wi th Comanche Peak, Indian Point 2, South Texas, Sun Deser t, W. B. McGuire Nuclear Station, Trojan Nuclear Power Plant, SNUPPS, and the Watts Bar Application. This inf ormation was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.
B/B-UFSAR 1.2-1 REVISION 15 - DECEMBER 2014 1.2 GENERAL PLANT DESCRIPTION 1.2.1 Site and Environment The characteristics of the sites and their environs have been investigated to establish bases for determining criteria for storm, flood, and earthquake protection and to evaluate the validity of calculational techniques for the control of routine and accidental releases of radioactive liquids and gases to the environment. Field programs to investigate geology and seismology are completed. Preoperational meteorological programs to provide onsite observations of wind speed and direction have continued since the spring of 1973 at Byron and since the fall of 1973 for Braidwood. Radiological studies of the site environs were initiated at least 18 months prior to commercial operation, with the objective of establishing background radiation levels.
B/B-UFSAR 1.3-2 REVISION 8 - DECEMBER 2000 1.3.2 Comparison of Final and Preliminary Information The Byron/Braidwood Power Plant design was s ubject to continuing review throughout the construction of the stations. The experience gained at Zion Stat ion and other PWRs was used to enhance equipment reli ability and perfor mance. Current design technology was used to upgrade earlier p lant design methods.
The geography, demography, meteorology, hydrology, geology, and seismology of the two plant sites are discussed in detail in Chapter 2.0.
No significant design ch anges have been made to the Byron Station or the Braidwood Stati on which have not been previously reported by amendment to the PS AR, except for the inclusion of 17 x 17
1.2.2 Nuclear Steam Supply System The nuclear steam supply system (NSSS) consists of a Westinghouse pressurized water reactor and supporting auxiliary systems.
Performance at the calculated steam flow of the NSSS at MUR conditions based on zero percent makeup is as follows:
a.
thermal output of NSSS (MWt) - 3659; b.
thermal output of reactor core (MWt) -3645; c.
steam flow from NSSS (lb/hr) - 16,347,514 for Unit 1/16,280,677 for Unit 2; d.
steam pressure at a steam generator outlet (psia) -
1020.8 for Unit 1 and 902 for Unit 2; e.
maximum moisture content (%) - 0.25%; and f.
feedwater temperature at steam generator inlet (F) -
446.5 for Unit 1 and 447.5 for Unit 2.
The NSSS consists of a reactor and closed reactor coolant loops connected in parallel to the reactor vessel, each loop containing a reactor coolant pump and a steam generator. The NSSS also contains an electrically heated pressurizer and certain auxiliary systems.


optimized fuel. Table 1.3-3 of the FSAR lis ted those significant changes reported since t he issuance of the B yron and Braidwood Stations Construction Pe rmits. This informati on was current at the time the Byron U nit 1 operating lice nse was granted and has not been included in the UFSAR.  
B/B-UFSAR 1.2-1a REVISION 7 - DECEMBER 1998 High pressure reactor coolant circulates through the reactor core to remove the heat generated by the nuclear reaction. The heated reactor coolant flows from the reactor vessel to the steam generators (via reactor coolant loop piping). The coolant gives up its heat to the feedwater in the steam generator to generate steam for the turbine generator. The cycle is completed when the reactor coolant is pumped back to the reactor vessel. The entire reactor coolant system is composed of leaktight components to contain the reactor coolant to the system.


Other changes included t he removal of the part length control rods (they are not n eeded to control X enon induced axial instabilities), the enla rgement of spent fue l capacity, the use of more corrosion-resist ant materials in the s team generators and moisture steam separator s, improved equipment packaging to do a reactor refueling in a shorter time peri od, an upgraded design for the reactor coolant pump seals, and replacement steam generators for Unit 1. These concepts are described in later chapters.  
B/B-UFSAR 1.2-2 REVISION 11 - DECEMBER 2006 The core is a multiregion type. All fuel assemblies are mechanically identical, although the fuel enrichment is not the same in all assemblies. In a typical initial core loading, three fuel enrichments are used in mechanically identical assemblies.
Fuel assemblies with the highest enrichments are placed in the core periphery, or outer region, and the two groups of lower enrichment fuel assemblies are arranged in a selected pattern in the central region. In subsequent refuelings, one third of the fuel is discharged and fresh fuel is loaded into the outer region of the core. The remaining fuel is arranged in the central two-thirds of the core in such a manner as to achieve optimum power distribution.
Rod cluster control assemblies are used for reactor control and consist of clusters of cylindrical absorber rods. The absorber rods move within guide tubes in certain fuel assemblies. Above the core, each cluster of absorber rods is attached to a spider connector and drive shaft, which is raised and lowered by a drive mechanism mounted on the reactor vessel head. The insertion of the rod cluster control assembly for a reactor trip is by gravity.
The reactor coolant pumps are Westinghouse vertical, single-stage, centrifugal pumps of the shaft-seal type.
The steam generators are B&W vertical U-tube units for Unit 1 and Westinghouse vertical U-tube units for Unit 2. All steam generators contain Inconel tubes. Integral moisture separation equipment reduces the moisture content of the steam.
The reactor coolant piping and all of the pressure-containing surfaces in contact with reactor water are stainless steel. The steam generator tubes and fuel cladding are Inconel and Zircaloy/ZIRLO, respectively. Reactor core internals, including control rod drive shafts, are primarily stainless steel.
An electrically heated pressurizer connected to one reactor coolant loop maintains reactor coolant system pressure during normal operation, limits pressure variations during plant load transients, and keeps system pressure within design limits during abnormal conditions.
Auxiliary system components are provided to charge makeup water into the reactor coolant system, purify reactor coolant, provide chemicals for corrosion inhibition and reactivity control, cool system components, remove decay heat, and provide for emergency safety injection.
1.2.3 Engineered Safety Features The engineered safety features provided for this plant have sufficient redundancy of components and power sources such that


====1.3.3 References====
B/B-UFSAR 1.2-3 REVISION 12 - DECEMBER 2008 under the conditions of a loss-of-coolant accident they can maintain the containment integrity and limit personnel exposure to less than 10 CFR 50.67 limits. The engineered safety features incorporated in the design of this plant and the functions they serve are summarized in the following.
: 1. Exelon Generatio n Company, "Byron/Br aidwood Stations Fire Protection Report in Response to Appendix A of BTP APCSB 9.5-1,"
a.
(current amendment).  
The emergency core cooling system injects borated water into the reactor coolant system if coolant is lost. This system limits damage to the core and limits the fission product contamination released into the containment following a postulated loss-of-coolant accident (LOCA).
b.
A steel lined, concrete containment vessel consists of a post-tensioned concrete cylindrical wall and shallow dome, and a conventionally reinforced concrete base. The containment forms a virtually leaktight barrier to prevent the escape of radioactivity.
c.
Reactor containment fan coolers reduce containment temperature and pressure following a postulated loss-of-coolant accident.
d.
A containment spray system is used to reduce containment pressure and to remove iodine and particulate fission products from the containment atmosphere in the event of a loss-of-coolant accident.
e.
The auxiliary feedwater system provides for heat removal from the reactor coolant system by providing makeup water to the steam generator under a variety of postulated conditions.
f.
A combustible gas control system is provided to ensure that the containment atmosphere is mixed following a loss-of-coolant accident. A mixed containment atmosphere prevents local accumulation of combustible or detonable gases that could threaten containment integrity or equipment operating in a local compartment.
1.2.4 Emergency Core Cooling System The emergency core cooling system (ECCS), with passive and active subsystems, is designed to inject borated water into the reactor coolant system (RCS) following a LOCA. This will provide cooling to limit core damage, metal-water reactions, and fission-product release. The ECCS provides long-term postaccident cooling of the core by drawing borated water from the containment sump.
1.2.5 Control and Instrumentation The reactor is controlled by a variety of reactivity coefficients (temperature, pressure, doppler) by control rod cluster motion which is required for load follow transients and for startup and shutdown, and by a soluble neutron absorber, i.e., boron in the


B/B-UFSAR 1.3-3 TABLE 1.3-1 PLANTS USING T HREE-BUTTRESS CO NTAINMENT DESIGN
B/B-UFSAR 1.2-4 REVISION 15 - DECEMBER 2014 form of boric acid which is adjusted in concentration to compensate for such effects as fuel consumption and accumulation of fission products.
1.2.6 Electrical System Each unit's main generator is an 1800-rpm, 3-phase, 60-cycle, hydrogen-innercooled unit with water-cooled stator windings and is rated at 1361 MVA at 75 psig gas pressure. Field excitation is provided by a direct shaft-driven brushless excitation system. Two one-half size main step-up transformers deliver power to the 345-kV switchyard.
The station's auxiliary power system consists of system and unit auxiliary transformers; 6900-V, 4160-V, and 480-V switchgear; 480-V motor control centers; 120-Vac instrument buses; and 250-Vdc and 125-Vdc buses.
Two diesel generators are provided for each unit and are available as onsite sources of power (in the event of complete loss of normal a-c power) for operating essential safeguard features. Each diesel generator is capable of supplying required electrical loads for a simultaneous LOCA and loss of offsite power to any one unit.
1.2.7 Turbine and Auxiliaries The turbine for each unit is a four-casing, tandem-compound, six-flow exhaust, 1800-rpm unit with 40-inch last-row blades. There are two combination moisture-separator/steam-reheater assemblies per unit. The turbine-generator for Units 1 have a MUR rating of 1268 MWe gross at 16,347,514 lb/hr steam flow with inlet steam conditions of 1001 psia, 0.36% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. The turbine-generators for Units 2 have a MUR rating of 1241 MWe gross at 16,280,677 lb/hr steam flow with inlet steam conditions of 882 psi, 0.34% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. There are seven stages of feedwater heating for all units.
The turbine is equipped with a redundant fault tolerant Westinghouse Ovation based distributed control system. All control algorithms and processes within the turbine control system are redundant and configured to allow unrestricted turbine operation. This system utilizes a fire-resistant hydraulic fluid to control throttle and governor valve positioning.


PLANT/UTILITY DATE OF OPERATION
B/B-UFSAR 1.2-4a REVISION 11 - DECEMBER 2006 The condenser is of the single-pass deaerating type. There are three parallel strings of feedwater heaters that utilize extraction steam from the low pressure turbines, two parallel strings of feedwater heaters that utilize extraction and exhaust steam from the high pressure turbine, four one-third-sized feedwater condensate and condensate booster pumps, and three one-half-sized feedwater and heater drain pumps. Heater drains from the three highest-pressure feedwater heaters are pumped into the feedwater system; drains from the four lowest-pressure heaters are cascaded to the condenser.


Arkansas Nuclear One
B/B-UFSAR 1.2-5 REVISION 14 - DECEMBER 2012 1.2.8 Fuel Handling System The reactor is refueled with equipment which handles the spent fuel under water from the entire time from leaving the reactor vessel until it is secured in a cask for shipment. Underwater transfer of spent fuel provides a transparent radiation shield and a reliable coolant for decay heat removal.
Fuel handling is performed in the refueling cavity which is flooded for refueling, and the fuel storage pool which is in the fuel building. The two areas are connected by a fuel transfer system which carries the fuel through an opening in the reactor containment.
Spent fuel is removed from the reactor vessel by a refueling machine, placed on the fuel transfer cart conveyor and transferred to the spent fuel storage pool. The fuel is removed from the transfer cart and placed into storage racks. After a suitable decay period, the fuel may be removed from storage and loaded into a shipping cask for removal from the plant.
Refer to Section 9.1.2.3.11 for a description of spent fuel storage and handling using Dry Cask Storage (DCS) system and the Independent Spent Fuel Storage Installation (ISFSI).
All important pumps, piping, and equipment are replicated and capable of being supplied from one of two independent ESF divisions.
1.2.9 Radioactive Waste Management System The radioactive waste system provides equipment necessary to collect, process, and prepare for the disposal of radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation or to transfer the wastes to a vendor-supplied radwaste system.
After collection, depending on chemical composition, liquid wastes may be demineralized and/or filtered. The treated water is discharged at concentrations within the limits of 10 CFR 20.
Sludges and spent demineralizer resins are processed by a vendor-supplied radwaste system for ultimate disposal in an authorized location.
Gaseous wastes are collected from the waste gas header.
Discharge of the gaseous wastes to the environment is controlled to ensure that the offsite dose is as low as reasonably achievable (ALARA).
1.2.10 Features of Special Interest The fundamental concept for the design and construction of the Byron/Braidwood Stations is one of commonality and duplication to the maximum extent permitted by site characteristics. For those features not dictated specifically by site characteristics, identical designs have been employed for the two stations. The concept has been extended to the point where the limiting (i.e.,


Arkansas Power & Light Co.
B/B-UFSAR 1.2-6 REVISION 9 - DECEMBER 2002 worst case) parameters of the sites are considered in the common design. An example of this is the use of the most restrictive site's seismic building response spectra for the design of systems and components in both plants.
5-21-74 Millstone-2
Common plans, drawings, and specifications have been issued for construction at the two sites. Design and construction management for both sites have been conducted by the same major organizations, using the same quality assurance and project management programs. This approach embraces the concept of standardization in nuclear power plant design and construction.
1.2.11 Structures The major structures include a separate and independent containment for each reactor, a common auxiliary building, a common turbine building, a common solid radwaste storage, and administration and service building. General layouts of the plant and interior components' arrangements are shown on Drawings M-5 through M-18 and M-20 and M-22 (Byron), and Drawings M-5 through M-20 and M-22 (Braidwood).
For purposes of design and analysis, structures are designated by Safety Category according to their relation to plant safety. The Safety Category definitions are as follows:
a.
Safety Category I - Those structures important to safety that must be designed to remain functional in the event of the safe shutdown earthquake (SSE) and other design-basis events (including tornado, probable maximum flood, operating basis earthquake (OBE), missile impact, or accident internal to the plant) are designated as Safety Category I.
b.
Safety Category II - Those structures which are not designated as Safety Category I are designated as Safety Category II.
The design criteria and analysis methods for these structures are discussed in Chapter 3.0.


Northeast Utilities 8-1-75 Rancho Seco Sacramento Municipal Utility District 8-16-74  Trojan Portland General Electric Co.
B/B-UFSAR 1.3-1 REVISION 8 - DECEMBER 2000 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs The design is conceptually similar to Exelon Generation Company's Zion Station. Differences in the design of the two plants have been allowed only (1) when dictated by the site characteristics, (2) when the change would result in significant safety improvement, simplification of construction or operation procedures, or cost savings; or (3) as required to comply with appropriate codes and standards, NRC criteria, regulatory guides, and regulations.
11-21-75
The nuclear steam supply system is similar to that of the Zion Station but has a slightly higher power rating. The reactor containments are of the same materials and size as those at the Zion Station, but each has only three buttresses, rather than six as used at Zion. The number of post-tensioning tendons is reduced, and the number of wires per tendon increased, from that used at Zion. The reduced number of buttresses allows for greater separation of penetration areas for redundant safety-related systems. Several plants on which this buttress design has been used are listed in Table 1.3-1.
The polar cranes in the reactor containment are mounted on the containment wall, rather than on the missile barrier as at Zion.
This allows use of a greater area for component laydown in the containment.
Two 100%-capacity containment spray systems are used, rather than the three systems used at Zion. Four containment fan coolers are used, rather than the five used at Zion. The emergency diesel-generator systems for each unit are entirely independent and use two 5500-kW diesel generators per unit. The arrangement of equipment in the common auxiliary building allows greater physical separation of redundant systems and their piping and cables than was possible at Zion.
The Byron Station uses natural draft cooling towers for heat rejection. Zion utilizes once-through cooling. Mechanical draft cooling towers are provided for essential service cooling at Byron.
The Braidwood Station uses a large man-made cooling pond for heat rejection. An auxiliary cooling pond, integral with the main pond, is provided for essential service cooling.
Table 1.3-2 of the FSAR provided the design comparison of the Byron/Braidwood nuclear steam supply system with Comanche Peak, Indian Point 2, South Texas, Sun Desert, W. B. McGuire Nuclear Station, Trojan Nuclear Power Plant, SNUPPS, and the Watts Bar Application. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.


J.M. Farley-1
B/B-UFSAR 1.3-2 REVISION 8 - DECEMBER 2000 1.3.2 Comparison of Final and Preliminary Information The Byron/Braidwood Power Plant design was subject to continuing review throughout the construction of the stations. The experience gained at Zion Station and other PWRs was used to enhance equipment reliability and performance. Current design technology was used to upgrade earlier plant design methods.
 
No significant design changes have been made to the Byron Station or the Braidwood Station which have not been previously reported by amendment to the PSAR, except for the inclusion of 17 x 17 optimized fuel. Table 1.3-3 of the FSAR listed those significant changes reported since the issuance of the Byron and Braidwood Stations Construction Permits. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.
Alabama Power Co.
Other changes included the removal of the part length control rods (they are not needed to control Xenon induced axial instabilities), the enlargement of spent fuel capacity, the use of more corrosion-resistant materials in the steam generators and moisture steam separators, improved equipment packaging to do a reactor refueling in a shorter time period, an upgraded design for the reactor coolant pump seals, and replacement steam generators for Unit 1. These concepts are described in later chapters.
6-25-77 B/B-UFSAR 1.4-1 REVISION 8 - DECEMBER 2000 1.4 IDENTIFICATION OF AG ENTS AND CONTRACTORS
1.3.3 References
: 1. Exelon Generation Company, "Byron/Braidwood Stations Fire Protection Report in Response to Appendix A of BTP APCSB 9.5-1,"
(current amendment).


====1.4.1 Licensee====
B/B-UFSAR 1.3-3 TABLE 1.3-1 PLANTS USING THREE-BUTTRESS CONTAINMENT DESIGN PLANT/UTILITY DATE OF OPERATION Arkansas Nuclear One Arkansas Power & Light Co.
5-21-74 Millstone-2 Northeast Utilities 8-1-75 Rancho Seco Sacramento Municipal Utility District 8-16-74 Trojan Portland General Electric Co.
11-21-75 J.M. Farley-1 Alabama Power Co.
6-25-77


Exelon Generation Company is the Licensee for the Byron Station, Units 1 and 2, which is located in Roc kvale Township, Ogle County, approximately 4 miles south of Byron
B/B-UFSAR 1.4-1 REVISION 8 - DECEMBER 2000 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 Licensee Exelon Generation Company is the Licensee for the Byron Station, Units 1 and 2, which is located in Rockvale Township, Ogle County, approximately 4 miles south of Byron, Illinois, and for Units 1 and 2 of the Braidwood Station, which is located in Reed Township, Will County, approximately 6 miles southwest of Wilmington, Illinois. The Licensee is responsible for the design, construction, and operation of the nuclear power plants.
, Illinois, and for Units 1 and 2 of the B raidwood Station, whic h is located in Reed Township, Will County, approximately 6 m iles southwest of Wilmington, Illinois.
Commonwealth Edison supplies electrical service to an area of 13,000 square miles with a population of approximately 8 million persons, located primarily in the northern third of Illinois.
The Licensee is r esponsible for the design, construction, and operation of the n uclear power plants.  
Dresden 1, Commonwealth Edison's first nuclear generating station, went into commercial service during August 1960, and has produced more than 10 billion kWh. Additional nuclear units in service or under construction are listed in Table 1.4-1.
1.4.2 Architect-Enqineer For the work covered by this application, Sargent & Lundy (S&L) has been retained as the design consultants. The Licensee has employed Sargent & Lundy for power plant design work for over 80 years.
Sargent & Lundy is an independent consulting engineering organization founded in Chicago, in 1891. For over three-quarters of a century, the firm has specialized exclusively in the design of generation, transmission, distribution, and utilization of steam and electric power and related facilities.
The firm has provided the complete engineering services for more than 600 turbine-generator units with a total capacity of 53,000,000 kW. Of this total, some 9,800,000 kW is in nuclear generating capacity. Table 1.4-2 lists the nuclear plants completed by or currently under design by Sargent & Lundy.
1.4.3 Reactor Designer Westinghouse has designed, developed, and manufactured nuclear power facilities since the 1950s, beginning with the world's first large central station nuclear power plant (Shippingport),
which started producing power in 1957. Completed or contracted


Commonwealth Edison su pplies electrical serv ice to an area of 13,000 square miles with a population of app roximately 8 million persons, located primarily in the northern thi rd of Illinois.  
B/B-UFSAR 1.4-2 REVISION 5 - DECEMBER 1994 commercial nuclear capacity totals were in excess of 98,000 MWe. Westinghouse pioneered new nuclear design concepts, such as chemical shim control of reactivity and the rod cluster control concept, throughout the last two decades. Westinghouse manufacturing facilities include the largest commercial nuclear fuel fabrication facility in the world and the world's most modern heat transfer equipment production facility, as well as other facilities producing nuclear steam supply system (NSSS) components. Table 1.4-3 lists all Westinghouse pressurized water reactor (PWR) plants to date, including those plants under construction or on order at the time of the Byron/Braidwood application.
The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute have contracted with Westinghouse for research into NSSS-related activities. Westinghouse experience was also utilized by the NRC and Metropolitan Edison immediately following the Three Mile Island Unit 2 accident and the corporation continues to participate with the Westinghouse Owner's Group of utilities in addressing the NRC action plan and other operations improvements.
1.4.4 Constructor Construction coordination of all activities at the site was under the supervision of the Commonwealth Edison's Station Construction Department. The department exercises site managerial functions as discussed in Chapter 17.0 of the UFSAR.
The Station Construction Department was the constructor for Zion Station. This department has coordinated the construction activities for almost all of Commonwealth Edison's existing power plants. It was also the construction coordinator for La Salle County Station.
1.4.5 Consultants and Service Organization 1.4.5.1 Security System - ETA The design of the physical security system and the administrative controls was performed by ETA, Inc.
ETA personnel have had varied and in-depth experience in the design, safety analysis, and environmental review of nuclear power plants and related facilities as well as in the management and organization of security systems. They are very familiar with the details of the current generation of light water reactors and, in particular, those critical areas and components of the plants which might be the most vulnerable to sabotage.
They are also familiar with the current regulations and guidelines of the NRC that define the required performance and objectives of a security system for licensed activities.  


Dresden 1, Commonwea lth Edison's first n uclear generating station, went into comme rcial service during A ugust 1960, and has produced more than 10 billion kWh.
B/B-UFSAR 1.4-3 1.4.5.2 Dames & Moore The independent consulting firm of Dames & Moore was employed to conduct studies relating to the geology, seismology, and groundwater hydrology at both sites. The firm also conducted preconstruction baseline studies, including wildlife surveys as well as soil and vegetation analyses.
Additional nucle ar units in service or under con struction are listed in Table 1.4-1.
Having performed environmental studies for approximately 30 nuclear power plant sites, Dames & Moore is a recognized authority in the field of environmental engineering of nuclear power plants.
 
1.4.5.3 HARZA Engineering HARZA was employed in the design of the water treatment facilities at both stations.
1.4.2  Architect-Enqineer
HARZA has been involved with a variety of technical studies for at least ten nuclear power stations. Among these studies have been facility design, review of design and structure, hydrology, and groundwater. In addition, HARZA Engineering has designed some of the largest hydroelectric projects in the world, including major concrete structures and earthfilled dams.
 
For the work covered by this application, Sa rgent & Lundy (S&L) has been retained as the design consultants. The Licensee has employed Sargent & Lundy for pow er plant design work for over 80 years. Sargent & Lundy is an independent cons ulting engineering organization founded in Chicago, in 1891.
For over three-quarters of a century, the firm has spec ialized exclusively in the design of generation, tr ansmission, dist ribution, and utilization of steam a nd electric power and related facilities.
The firm has provided the comple te engineering s ervices for more than 600 turbine-generator u nits with a total capacity of 53,000,000 kW. Of t his total, some 9,80 0,000 kW is in nuclear generating capacity.
Table 1.4-2 lists the nuclear plants completed by or curr ently under design by Sargent & Lundy.
1.4.3  Reactor Designer
 
Westinghouse has designed, dev eloped, and manu factured nuclear power facilities since t he 1950s, beginning with the world's first large central stat ion nuclear power pl ant (Shippingport),
which started producing power in 1957. Comple ted or contracted
 
B/B-UFSAR 1.4-2 REVISION 5 - DECEMBER 1994 commercial nuclear cap acity totals were in excess of 98,000 MWe. Westinghouse p ioneered new nuclear design concepts, such as chemical shim control of reac tivity and the rod cluster control concept, through out the last two decad es. Westinghouse manufacturing facilities include the largest com mercial nuclear fuel fabrication facility in the world and t he world's most modern heat transfer e quipment production facility, as well as other facilities producing nucle ar steam supply system (NSSS) components. Table 1
.4-3 lists all Wes tinghouse pressurized water reactor (PWR) pl ants to date, incl uding those plants under construction or on ord er at the time of the Byron/Braidwood application.
The U.S. Nuclear Regulatory Commissi on (NRC) and the Electric Power Research I nstitute have contracted with Westinghouse for research into NSSS-related activ ities. Westinghouse experience was also utilized by the NRC and Metropolitan Edison immediately following the Three Mile Island Unit 2 accident and the corporation continues to participate wit h the Westinghouse Owner's Group of utilities in ad dressing the NRC action plan and other operations improvements.
 
====1.4.4 Constructor====
Construction coordination of all activities at the site was under the supervision of the C ommonwealth Edison's Station Construction Department. Th e department exercises site managerial functions as discussed in Chapter 1 7.0 of the UFSAR.
The Station Construction Departm ent was the constructor for Zion Station. This departm ent has coordinated the construction activities for almost all of Commonwealth Ed ison's existing power plants. It was also the construction coordi nator for La Salle County Station.
 
1.4.5  Consultants and Service Organization 1.4.5.1 Security System - ETA
 
The design of the physical secur ity system and the administrative controls was perform ed by ETA, Inc.
ETA personnel have had varied and in-depth experience in the design, safety analysis, and environmental r eview of nuclear power plants and related facilit ies as well as in the management and organization of security sys tems. They are very familiar with the details of the current generation of light water reactors and, in particular, those critical areas and components of the plants wh ich might be the most vulnerable to sabotage.
They are also famili ar with the current regulations and guidelines of the NRC that define the required performance and objectives of a security system for licensed activities.
 
B/B-UFSAR 1.4-3 1.4.5.2 Dames & Moore The independent consulti ng firm of Dames &
Moore was employed to conduct studies relating to the geo logy, seismology, and groundwater hydrology at both sites. The firm also conducted preconstruction baseline studies, including wi ldlife surveys as well as soil and vegetation analyses.
Having performed envir onmental studies for approximately 30 nuclear power plant si tes, Dames & Moore is a recognized authority in the field of en vironmental engine ering of nuclear power plants.
1.4.5.3 HARZA Engineering
 
HARZA was employed in the design of the water treatment facilities at both stations.
HARZA has been involved with a variety of technical studies for at least ten nuclear power sta tions. Among these studies have been facility design, review of design and structure, hydrology, and groundwater. In addition, HARZA Eng ineering has designed some of the largest hy droelectric projects in the world, including major concrete struc tures and earthfilled dams.
1.4.5.4 Murray and Trettel, Inc.
1.4.5.4 Murray and Trettel, Inc.
Murray and Trettel (M&T) is an environmental consulting firm which, since 1960, has provided significant meteorological input to both preoperational and operational phases of meteorological programs for nuclear power stations. M&T has also provided meteorological input to a wide variety of air pollution and environmental problems as well as allied control technique programs.
Murray & Trettel provided meteorological data for both stations by implementation of an onsite measurement program incorporating a tower for elevation measurements.
1.4.5.5 Shirmer Engineering Corporation Shirmer Engineering is a firm of consulting fire protection engineers. The firm has done work on 17 Department of Energy nuclear fuel production and laboratory facilities, as well as for numerous nuclear power stations for Sargent & Lundy.
Shirmer Engineering has also performed services for many fossil units.
Shirmer Engineering provided evaluation of the fire protection systems at both stations and assisted in the preparation of the Byron/Braidwood Fire Protection Report.


Murray and Trettel (
B/B-UFSAR 1.4-4 1.4.5.6 Hyla S. Napadensky Ms. Napadensky was retained to help evaluate the probability of an accidental explosion occurring on a train carrying explosives in the vicinity of the Braidwood Station.
M&T) is an environmental consulting firm which, since 1960, has provided signif icant meteorological input to both preoperational and operational phases of meteorological progr ams for nuclear power stat ions. M&T has also provided meteorological input to a wide variety of air pollution and environmental prob lems as well as allied control technique programs.
Ms. Napadensky is the Manager of Fire Safety Research at the IIT Research Institute of the Illinois Institute of Technology.
 
Ms. Napadensky has directed analytical and experimental research in the areas of explosion effects, hazards and risk analysis, safety of chemical systems, explosives and propellant sensitivity, and initiation mechanisms during her 17 years with IIT Research Institute.
Murray & Trettel provided meteor ological data for both stations by implementation of an onsite measurement p rogram incorporating a tower for elev ation measurements.
1.4.5.7 NALCO Chemical Company The NALCO Chemical Company (formerly Industrial Bio-Test, Inc.)
1.4.5.5 Shirmer Engi neering Corporation
consisted of two divisions, Industrial Bio-Test Laboratories, and NALCO Environmental Sciences, which conduct studies relating to toxicology and ecological sciences, respectively. The Environmental Science Division includes seven subdivisions: (1) aquatic biology, (2) fisheries and field operations, (3) water and wastewater chemistry, (4) radiochemistry, (5) air sciences and data processing, (6) land and plant sciences, and (7) environmental physiology.
 
As a technical consultant on the Braidwood project, the NALCO Chemical Company provided a clam bed mapping survey in the area of the station's intake and discharge structures located on the Kankakee River.
Shirmer Engineering is a firm of consulting fire protection engineers. The firm has done work on 17 Department of Energy nuclear fuel production and laboratory facilit ies, as well as for numerous nuclear power stations for Sargent & Lundy.
1.4.5.8 Westinghouse Environmental Systems Department (WESD)
Shirmer Engineering has also performed servi ces for many fossil units. Shirmer Engineering prov ided evaluation of the fire protection systems at both stations and assisted in the preparation of the Byron/Braidwood Fire Protection Report.
WESD, established as a department of the Westinghouse Power Systems Company in 1969, consisted of environmental scientists and engineers experienced in the areas of aquatic and terrestrial biology and ecology; geology; limnology; environmental chemistry and physics; physical oceanography, meteorology and climatology, radiology, public health aspects of pollutant emissions, and systems engineering and integration.
 
WESD conducts broad environmental surveys, environmental program planning and data interpretation, and provides recommended action programs for meeting federal, state, and local environmental quality regulations. As a technical consultant on the Braidwood project, WESD staff biologists conducted a 2-year baseline study of the Braidwood Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Kankakee River in the vicinity of the site were determined, and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation in the biotic communities of the site were predicted.  
B/B-UFSAR 1.4-4 1.4.5.6 Hyla S. Napadensky Ms. Napadensky was ret ained to help evaluate the probability of an accidental explosion occurring on a train carrying explosives in the vicinity of t he Braidwood Station.
Ms. Napadensky is the Manager of Fire Safety Research at the IIT Research Institute of the Illinois Insti tute of Technology.
Ms. Napadensky has dir ected analytical and e xperimental research in the areas of explosion effects, haz ards and r isk analysis, safety of chemical systems, explosives and propellant sensitivity, and initiat ion mechanisms during her 17 years with IIT Research Institute.  
 
1.4.5.7 NALCO Ch emical Company
 
The NALCO Chemical Com pany (formerly Industr ial Bio-Test, Inc.)
consisted of two divisions, Indu strial Bio-Test Labo ratories, and NALCO Environmental Scie nces, which conduct st udies relating to toxicology and ecolo gical sciences, re spectively. The Environmental Science Di vision includes seven subdivisions: (1) aquatic biology, (2) fis heries and field opera tions, (3) water and wastewater chemistry, (4) radiochemistry, (5) air sciences and data processing, (6) land and plant sciences, and (7) environmental physiology.
As a technical consult ant on the Braidwo od project, the NALCO Chemical Company provided a clam bed mapping sur vey in the area of the station's intake and discharge struct ures located on the Kankakee River.
1.4.5.8 Westinghouse E nvironmental Syste ms Department (WESD)
 
WESD, established as a department of t he Westinghouse Power Systems Company in 1969, consisted of environmental scientists and engineers experienced in the areas of aqua tic and terrestrial biology and ecology; g eology; limnology; env ironmental chemistry and physics; physical oceanography, meteorol ogy and climatology, radiology, public health aspec ts of pollutant emissions, and systems engineering and integration.  


WESD conducts broad en vironmental surveys, environmental program planning and data interpretation, and provides recom mended action programs for meeting federal, state, and local environmental quality regulations. As a technical consult ant on the Braidwood project, WESD staff biol ogists conducted a 2-y ear baseline study of the Braidwood Stati on site. Distribution s of phytoplankton, zooplankton, periphyton, benthos, fi sh, fish eggs and larvae, and water chemistry in the Kankakee Rive r in the vicinity of the site were determined, and quantitative data on terr estrial flora and fauna were collected.
The impacts of plant construction and operation in the biotic communit ies of the site were predicted.
B/B-UFSAR 1.4-5 1.4.5.9 Illinois Natural History Survey (INHS)
B/B-UFSAR 1.4-5 1.4.5.9 Illinois Natural History Survey (INHS)
The Illinois Natural History Survey (INH S), which has its beginnings almost 12 0 years ago, is a di vision of the State Department of Registra tion and Education and provides services to farmers, homeowners, sportsmen, and all other citizens of Illinois as well as to industries.
The Illinois Natural History Survey (INHS), which has its beginnings almost 120 years ago, is a division of the State Department of Registration and Education and provides services to farmers, homeowners, sportsmen, and all other citizens of Illinois as well as to industries. INHS cooperates in biological research with the Illinois Department of Agriculture, Conservation, and Public Health; the University of Illinois, Southern Illinois University, and other educational institutions; various research branches of the federal government; and other agricultural, conservation, municipal, and business organizations throughout the state.
INHS cooperates in biological research with the Illinois Department of Agriculture, Conservation, and Public Health; the University of Illinois, Southern Illinois University, and other educational institutions; various research branches of t he federal government; and other agricultural, conservati on, municipal, and bus iness organizations throughout t he state.
INHS aquatic biologists were involved in a 4-year study of the Kankakee River and Horse Creek near Custer Park, Illinois. The purpose of the study is to obtain biological, physical, and chemical data which will be used to evaluate any effects of the construction and operation of the Braidwood Station and its associated cooling lake on the biota and water quality of the Kankakee River and Horse Creek. The station's cooling pond will use the Kankakee River as a source of water for both intake and discharge purposes.
INHS aquatic biologists were involved in a 4
1.4.5.10 NUS Corporation NUS Corporation is a consulting engineering, research, and testing firm specializing in environmental and energy systems engineering, systems analysis, design engineering, management consulting, and training programs related to these areas. NUS has provided advice and professional guidance to utility, industrial, and government clients throughout the United States and in a number of foreign countries.
-year study of the Kankakee River and Hor se Creek near Custer P ark, Illinois. The purpose of the s tudy is to obtain bi ological, physical, and chemical data which will be used to evaluate any effects of the construction and operation of the Braidwood Station and its associated cooling lake on the biota and water quality of the Kankakee River and Hor se Creek. The sta tion's cooling pond will  
As a technical consultant on the Braidwood project, NUS was involved in a study to determine the adequacy of the station's ultimate heat sink.
1.4.5.11 Eberline Instrument Corporation (EIC)
Eberline Instrument Corporation (EIC) has provided radiation measurement equipment, comprehensive radiation protection services, and analytical laboratory services to the nuclear industry since 1953.
As a technical consultant on the Byron/Braidwood projects EIC performed preoperational environmental radiological baseline studies on and around the site.
1.4.5.12 Meteorology Research, Inc. (MRI)
Meteorology Research, Inc. (MRI) is an environmental consulting firm which, since 1951, has provided meteorological and air


use the Kankakee River as a source of water for both intake and discharge purposes.  
B/B-UFSAR 1.4-6 REVISION 1 - DECEMBER 1989 quality instruments and services to all aspects of industry in the solution of weather-related problems. These range from environmental impact assessments of existing or proposed airports and other major developments to problems of warehousing and marketing seasonal consumer goods. Of particular interest is the influence the local topography has on temperatures and winds.
MRI provided meteorological data from 1973 through mid-1975 for Byron and Braidwood Stations by implementation of an onsite meteorological measurement program.
1.4.5.13 Illinois State Museum (ISM)
The Illinois State Museum conducts archaeological investigations throughout the state of Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing and excavating, and laboratory analyses of datifacts, pollen, and soils.
As a technical consultant on the Braidwood project, ISM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Braidwood Station and pipeline corridor property.
1.4.5.14 Equitable Environmental Health, Inc. (EEH)
Equitable Environmental Health, Inc. (EEH), successor to Environmental Analysts, Inc./Tabershaw-Cooper Associated, Inc.,
is a multidisciplinary firm that offers the consulting services of medical professionals, scientists, engineers, economists, and technical support personnel in all areas of environmental health and economics.
EEH staff biologists conducted a 2-year baseline study of the Byron Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Rock River in the vicinity of the site were determined and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation on the biotic communities of the site were predicted, and data were provided for a benefit-cost analysis of the project.
1.4.5.15 Espey, Huston & Associates, Inc. (EH & A)
Espey, Huston & Associates, Inc. (EH & A) is a consulting firm addressing the environmental problems associated with industrial and urban development. EH & A professionals cover a broad range of expertise including civil engineering, environmental engineering, mathematics and computer science, and all phases of aquatic, estuarine, and terrestrial ecology.  


1.4.5.10 NUS Corporation
B/B-UFSAR 1.4-7 REVISION 1 - DECEMBER 1989 As a technical consultant on the Byron project, EH & A conducted the construction phase terrestrial and aquatic monitoring programs.
1.4.5.16 University of Wisconsin-Milwaukee (UWM)
The University of Wisconsin-Milwaukee under Dr. Elizabeth Benchley of the Dept. of Anthropology, conducts archaeological investigations throughout Wisconsin and northern Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing, and lab analysis of datifacts, pollen, and soils.
As a technical consultant on the Byron project, UWM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Byron Station and pipeline corridor property. Also, UWM conducted archaeological investigations on the Byron transmission line right-of-ways.
1.4.5.17 Aero-Metric Engineering, Inc. (AME)
Aero-Metric Engineering, Inc., founded in 1969, is based in Sheboygan, Wisconsin. The staff was made up of over 50 technical photogrammetric personnel, many having professional engineer and/or survey registration. AME's capabilities allow for a complete range of precision photogrammetric services, including aerial photography, mapping, and multiple survey skills.
As a technical consultant on the Byron project, AME will be providing annual aerial infra-red photographs.
1.4.5.18 Iowa Institute of Hydraulic Research The Iowa Institute of Hydraulic Research, formally organized in 1931, is a Division of the University of Iowa's College of Engineering. The Institute staff exceeded 80 in number and was comprised of a professional staff with Ph.Ds in the areas of Civil Engineering, Mechanical Engineering, Physics, Mechanics and Hydraulics, and Aeronautical Engineering, with most of these personnel holding joint academic appointments in the College of Engineering's Division of Energy Engineering. The Institute of Hydraulic Research conducts programs of fundamental research and advanced design and analysis in the areas of environmental pollution, bioengineering, naval hydrodynamics, river mechanics, ice hydraulics, hydrology, water resources, hydraulic structures, fluid mechanics, advanced instrumentation and data-handling techniques for fluids research, and mathematical modeling of watersheds and hydrology.


NUS Corporation is a consulting engineer ing, research, and testing firm specializ ing in environment al and energ y systems engineering, systems analysis, design engine ering, management consulting, and training pro grams related to these areas. NUS has provided advice and professional g uidance to utility, industrial, and governme nt clients throughou t the United States and in a number of f oreign countries.  
B/B-UFSAR 1.4-8 REVISION 9 - DECEMBER 2002 As a technical consultant on the Braidwood project, the Institute conducted a thermal evaluation to determine the adequacy of the ultimate heat sink.
1.4.5.19 Babcock and Wilcox International (B&W)
B&W is located in Cambridge, Ontario, Canada. B&W has fabricated fossil-fueled boiler components for over 100 years and has fabricated nuclear system components since the late 1950's. B&W has supplied replacement steam generators for Byron Unit 1 and Braidwood Unit 1.
1.4.5.20 Framatome Technologies, Incorporated (FTI)
FTI is located in Lynchburg, Virginia and has been providing services to the electric power industry for over four decades.
FTI engineering services include the necessary expertise, experience, and NRC-approved computer codes and methodologies to support the transient analysis of the Unit steam generators.
1.4.5.21 Stone & Webster Engineers and Constructors, Inc. (S&W)
S&W is located in Boston, Massachusetts and has been providing services to the electric power industry for over 100 years. S&W has provided balance-of-plant design-engineering support services in support of the power uprate of the Byron and Braidwood units.  


As a technical consult ant on the Braidwo od project, NUS was involved in a study to determine the adequacy of the station's ultimate heat sink.  
B/B-UFSAR 1.4-9 REVISION 8 - DECEMBER 2000 TABLE 1.4-1 EXELON GENERATION COMPANY'S NUCLEAR POWER PLANTS IN SERVICE OR UNDER CONSTRUCTION UNIT NOMINAL GROSS1 RATING (MWe)
SCHEDULED COMMERCIAL SERVICE DATE Dresden 1 210 1960 Dresden 2 850 1972 Dresden 3 850 1972 Quad-Cities 1 850 1972 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 La Salle 1 1122 1978 La Salle 2 1122 1979 Byron 1 1175 1985 Byron 2 1175 1987 Braidwood 1 1175 1988 Braidwood 2 1175 1988 1Note that this is a gross rating, not a net rating.  


1.4.5.11 Eberline Instrum ent Corporation (EIC)
B/B-UFSAR 1.4-10 REVISION 1 - DECEMBER 1989 TABLE 1.4-2 NUCLEAR POWER PLANTS COMPLETED OR CURRENTLY UNDER DESIGN BY SARGENT & LUNDY UNIT NOMINAL GROSS2 RATING (MWe)
YEAR OF POWER OPERATION EBWR 5
1956 Elk River 22 1962 La Crosse 60 1967 SEFOR 20 (MWt) 1969 Dresden 2 850 1969 Dresden 3 850 1971 Quad-Cities 1 850 1971 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 Fort St. Vrain, Unit 1 330 1973 Enrico Fermi, Unit 2 1200 1988 La Salle County Station, Unit 1 1122 1979 La Salle County Station, Unit 2 1122 1980 Byron Station, Unit 1 1175 1985 Byron Station, Unit 2 1175 1987 Braidwood Station, Unit 1 1175 1988 Braidwood Station, Unit 2 1175 1988 Clinton Power Station, Unit 1 992 1981 Kaiseraugst 992 1982 2Note that this is a gross rating, not a net rating.


Eberline Instrument Corp oration (EIC) has pr ovided radiation measurement equipment, comprehensive rad iation protection services, and analytical labor atory services to the nuclear industry since 1953.
B/B-UFSAR 1.4-11 TABLE 1.4-3 WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTS PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Shippingport Duquesne Light Company; Energy Research & Development Administration Pennsylvania 1957 90 4
As a technical consult ant on the Byron/Braidwood projects EIC performed preoperational envir onmental radiolo gical baseline studies on and a round the site.
Yankee-Rowe Yankee Atomic Electric Company Massachusetts 1961 175 4
1.4.5.12 Meteorology Re search, Inc. (MRI)
Trio Vercellese (Enrico Fermi)
Meteorology Research, In
Ente Nazionale per L'Energia Elettrica (ENEL)
: c. (MRI) is an environ mental consulting firm which, since 19 51, has provided m eteorological and air
Italy 1965 260 4
Chooz (Ardennes)
Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA)
France 1967 305 4
San Onofre Unit 1 Southern California Edison Co.;
San Diego Gas and Electric Co.
California 1968 450 3
Haddam Neck (Connecticut Yankee)
Connecticut Yankee Atomic Power Company Connecticut 1968 575 4
Jose Cabrera-Zorita Union Electrica, S.A.
Spain 1969 153 1
Beznau Unit 1 Nordostschweizerische Krafwerke AG (NOK)
Switzerland 1969 350 2
Robert Emmett Ginna Rochester Gas and Electric Corporation New York 1970 490 2
Mihama Unit 1 The Kansai Electric Power Company, Inc.
Japan 1970 320 2
Point Beach Unit 1 Wisconsin Electric Power Co.;
Wisconsin Michigan Power Co.
Wisconsin 1970 497 2
H. B. Robinson Unit 2 Carolina Power and Light Co.
South Carolina 1971 707 3


B/B-UFSAR 1.4-6 REVISION 1 - DECEMBER 1989 quality instruments and services to all aspects of industry in the solution of weathe r-related problems.
B/B-UFSAR 1.4-12 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
These range from environmental impact a ssessments of existing or proposed airports and other major deve lopments to proble ms of warehousing and marketing seasonal con sumer goods. Of particu lar interest is the influence the local to pography has on temper atures and winds.
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beznau Unit 2 Nordostschweizerische Kraftwerke AG (NOK)
MRI provided meteorolo gical data from 1973 t hrough mid-1975 for Byron and Braidwood St ations by implementati on of an onsite meteorological measu rement program.
Switzerland 1972 350 2
1.4.5.13 Illinois Sta te Museum (ISM)
Point Beach Unit 2 Wisconsin Electric Power Co.;
Wisconsin Michigan Power Co.
Wisconsin 1972 497 2
Surry Unit 1 Virginia Electric and Power Co.
Virginia 1972 822 3
Turkey Point Unit 3 Florida Power and Light Co.
Florida 1972 745 3
Indian Point Unit 2 Consolidated Edison Company of New York, Inc.
New York 1973 873 4
Prairie Island Unit 1 Northern States Power Company Minnesota 1973 530 2
Turkey Point Unit 4 Florida Power and Light Co.
Florida 1973 745 3
Surry Unit 2 Virginia Electric and Power Co.
Virginia 1973 822 3
Zion Unit 1 Exelon Generation Company Illinois 1973 1050 4
Kewaunee Wisconsin Public Service Corp.;
Wisconsin Power and Light Co.;
Madison Gas and Electric Co.
Wisconsin 1974 560 2
Prairie Island Unit 2 Northern States Power Company Minnesota 1974 530 2
Takahama Unit 1 The Kansai Electric Power Company, Inc.
Japan 1974 781 3
Zion Unit 2 Exelon Generation Company Illinois 1974 1050 4


The Illinois State Mus eum conducts archaeolo gical investigations throughout the s tate of Illinois. As a member of the Illinois Archaeological Survey, t hey have the experti se and services to perform contract archaeo logical work. Their studies included a pedestrian reconnaissance surv ey, subsurface testing and excavating, and laborato ry analyses of datifac ts, pollen, and soils.
B/B-UFSAR 1.4-13 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
As a technical consult ant on the Braid wood project, ISM identified and made re commendations which Commonwealth Edison acted upon to aid in preserv ing the archaeolog ical sites on Braidwood Station and pi peline corridor property.  
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Doel Unit 1 Indivision Doel Belgium 1975 390 2
Doel Unit 2 Indivision Doel Belgium 1975 390 2
Donald C. Cook Unit 1 Indiana and Michigan Electric Company (AEP)
Michigan 1975 1060 4
Ringhals Unit 2 Statens Vattenfallsverk (SSPB)
Sweden 1975 822 3
Almaraz Unit 1 Unit Electrica, S.A.;
Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A.
Spain 1976 902 3
Beaver Valley Unit 1 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Company Pennsylvania 1976 852 3
Diablo Canyon Unit 1 Pacific Gas and Electric Co.
California 1976 1084 4
Indian Point Unit 3 Consolidated Edison Company of New York, Inc.
New York 1976 965 4
Lemoniz Unit 1 Iberduero, S.A.
Spain 1976 902 3
Salem Unit 1 Public Service Electric and Gas Company; Exelon Generation Company; Atlantic City Electric Co.;
Delmarva Power and Light Co.
New Jersey 1976 1090 4


1.4.5.14 Equitable Environmen tal Health, Inc. (EEH)
B/B-UFSAR 1.4-14 TABLE 1.4-3 (Cont'd)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Trojan Portland General Electric Co.;
Eugene Water and Electric Board; Pacific Power and Light Company Oregon 1976 1130 4
Almaraz Unit 2 Union Electrica, S.A.;
Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A.
Spain 1977 902 3
Asco Unit 1 Fuerzas Electricas de Cataluna, S.A. (FESCA)
Spain 1977 902 3
Diablo Canyon Unit 2 Pacific Gas and Electric Co.
California 1977 1106 4
Joseph M. Farley Unit 1 Alabama Power Company Alabama 1977 829 3
Ko-Ri Unit 1 Korea Electric Power Co., Ltd.
Korea 1977 564 2
North Anna Unit 1 Virginia Electric and Power Co.
Virginia 1977 898 3
North Anna Unit 2 Virginia Electric and Power Co.
Virginia 1977 898 3
Ohi Unit 1 The Kansai Electric Power Co., Inc.
Japan 1977 1122 4
Ohi Unit 2 The Kansai Electric Power Co., Inc.
Japan 1977 1122 4
Ringhals Unit 3 Statens Vattenfallsvert (SSPB)
Sweden 1977 900 3
Sequoyah Unit 1 Tennessee Valley Authority Tennessee 1977 1148 4
Angra dos Reis Unit 1 Furnas-Centrais Electricas, S.A.
Brazil 1978 626 2


Equitable Environmen tal Health, Inc. (
B/B-UFSAR 1.4-15 TABLE 1.4-3 (Cont'd)
EEH), successor to Environmental Analysts, Inc./Tabershaw-Cooper Associated, Inc.,
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Asco Unit 2 Fuerzas Electricas de Cataluna, S.A. (FESCA);
is a multidisciplinary firm that offers the consulting services of medical professionals, sc ientists, engineer s, economists, and technical support pe rsonnel in all areas of environmental health and e conomics.
Empresa Nacional Hidroelectrica del Ribagorzana, S.A. (ENHER);
Fuerzas Hidroelectricas del Segre, S.A.;
Hidroelectrica de Cataluna, S.A.
Spain 1978 902 3
Donald C. Cook Unit 2 Indiana and Michigan Electric Company (AEP)
Michigan 1978 1060 4
Lemoniz Unit 2 Iberduero, S.A.
Spain 1978 902 3
Sequoyah Unit 2 Tennessee Valley Authority Tennessee 1978 1148 4
Watts Bar Unit 1 Tennessee Valley Authority Tennessee 1978 1177 4
William B. McGuire Unit 1 Duke Power Company North Carolina 1978 1180 4
Joseph M. Farley Unit 2 Alabama Power Company Alabama 1979 829 3
Krsko Savske Elektrarne, Ljubljana, Slovenia, Elektroprivreda, Zagreb, Croatia Yugoslavia 1979 615 2
Ringhals Unit 4 Statens Vattenfallsvert (SSPD)
Sweden 1979 900 3


EEH staff biologists conducted a 2-year base line study of the Byron Station site. Dis tributions of phytopla nkton, zooplankton, periphyton, benthos, f ish, fish eggs and larvae, and water chemistry in the Rock River in t he vicinity of t he site were determined and q uantitative data on terrestr ial flora and fauna were collected. The impacts of plant constr uction and operation on the biotic communities of the site were predicted, and data were provided for a benefit-cost analysis of the project.
B/B-UFSAR 1.4-16 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
1.4.5.15 Espey, Huston & Asso ciates, Inc. (EH & A)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Salem Unit 2 Public Service Electric and Gas Company; Exelon Generation Company Atlantic City Electric Co.;
Delmarva Power and Light Co.
New Jersey 1979 1115 4
Virgil C. Summer South Carolina Electric and Gas Company South Carolina 1979 900 3
Watts Bar Unit 2 Tennessee Valley Authority Tennessee 1979 1177 4
William B. McGuire Unit 2 Duke Power Company North Carolina 1979 1180 4
Byron Unit 1 Exelon Generation Company Illinois 1981 1120 4
Comanche Peak Unit 1 Texas Utilities Generating Co.
Texas 1980 1150 4
Seabrook Unit 1 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1980 1200 4
South Texas Project Unit 1 Houston Lighting and Power Co.;
Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Texas 1980 1250 4


Espey, Huston & Associat es, Inc. (EH & A) is a consulting firm addressing the e nvironmental problems as sociated with industrial and urban development.
B/B-UFSAR 1.4-17 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
EH & A professionals c over a broad range of expertise includi ng civil engineeri ng, environmental engineering, mathematics and com puter science, and all phases of aquatic, estuarine, and terrestrial ecology.
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beaver Valley Unit 2 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Co.;
B/B-UFSAR 1.4-7 REVISION 1 - DECEMBER 1989 As a technical consultant on the Byron project, EH & A conducted the construction p hase terrestrial and aquatic monitoring programs.
Cleveland Electric Illuminating Company; Toledo Edison Company Pennsylvania 1981 852 3
1.4.5.16 University of Wi sconsin-Milwaukee (UWM)
Braidwood Unit 1 Exelon Generation Company Illinois 1981 1120 4
The University of Wi sconsin-Milwaukee un der Dr. Elizabeth Benchley of the Dept. of Anthrop ology, conducts archaeological investigations throughout Wisconsin and nort hern Illinois. As a member of the Illinois Archaeo logical Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestri an reconnaissance survey, subsurface t esting, and lab anal ysis of datifacts, pollen, and soils.
Callaway Unit 1 SNUPPS - Union Electric Co.
As a technical consult ant on the Byron p roject, UWM identified and made recommendations which C ommonwealth Edison acted upon to aid in preserving the arc haeological sites on Byron Station and pipeline corrido r property. Also, UWM conducted archaeological investiga tions on the Byron transmission line right-of-ways.
Missouri 1981 1150 4
1.4.5.17 Aero-Metric En gineering, Inc. (AME)
Catawba Unit 1 Duke Power Company South Carolina 1981 1153 4
Aero-Metric Engineering, Inc., f ounded in 1969, is based in Sheboygan, Wisconsin.
Jamesport Unit 1 Long Island Lighting Company New York 1981 1150 4
The staff was made up of over 50 technical photogrammetric person nel, many having professional engineer and/or survey registration. AM E's capabilities allow for a complete range of precision phot ogrammetric services, including aerial photography, mapping, and m ultiple survey skills.
Ko-Ri Unit 2 Korea Electric Power Co., Ltd.
As a technical consult ant on the Byron project, AME will be providing annual aerial infra-red photographs.
Korea 1981 605 2
1.4.5.18 Iowa Institute of Hydraulic Research The Iowa Institute of Hy draulic Research, form ally organized in 1931, is a Division of the University of Iowa's College of Engineering. The Institute staff exceeded 80 in number and was comprised of a profess ional staff with Ph.Ds in the areas of Civil Engineering, Mec hanical Engineering, P hysics, Mechanics and Hydraulics, and Aero nautical Engineering, with most of these personnel holding joint academic a ppointments in the College of Engineering
NORCO Puerto Rico Water Resources Authority Puerto Rico 583 2
's Division of Ene rgy Engineering. The Institute of Hydraul ic Research conducts programs of fundamental research and advanced design and analysis in the areas of environmental pollution, bioeng ineering, naval hydrodynamics, river mechanics, ice hydraulics, hydrology, water resources, hydraulic struc tures, fluid m echanics, advanced instrumentation and dat a-handling techniques for fluids research, and m athematical modeling of watersheds and hydrology.
Braidwood Unit 2 Exelon Generation Company Illinois 1982 1120 4
Byron Unit 2 Exelon Generation Company Illinois 1982 1120 4
Catawba Unit 2 Duke Power Company South Carolina 1982 1153 4
Comanche Peak Unit 2 Texas Utilities Generating Co.
Texas 1982 1150 4
Marble Hill Unit 1 Public Service Company of Indiana, Inc.;
Northern Indiana Public Service Company Indiana 1982 1150 4


B/B-UFSAR 1.4-8 REVISION 9 - DECEMBER 2002 As a technical consult ant on the Braid wood project, the Institute conducted a thermal evaluation to determine the adequacy of the ulti mate heat sink.
B/B-UFSAR 1.4-18 TABLE 1.4-3 (Cont'd)
1.4.5.19 Babcock and Wilc ox International (B&W)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Millstone Unit 3 Northeast Nuclear Energy Co.
B&W is located in Cambri dge, Ontario, Canada.
Connecticut 1982 1156 4
B&W has fabricated fossil-fueled boiler c omponents for over 100 years and has fabricated nuclear system compon ents since the l ate 1950's. B&W has supplied replaceme nt steam generators fo r Byron Unit 1 and Braidwood Unit 1.
Seabrook Unit 3 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1982 1200 4
South Texas Project Unit 2 Houston Lighting and Power Co.;
Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Texas 1982 1250 4
Taiwan Unit 5 Taiwan Power Company Taiwan 1982 950 3
Wolf Creek Unit 1 SNUPPS - Kansas Gas and Electric Company; Kansas City Power and Light Company Kansas 1982 1150 4
Alvin W. Vogtle Unit 1 Georgia Power Company Georgia 1983 1113 4
Callaway Unit 2 SNUPPS - Union Electric Company Missouri 1983 1150 4
NEP-1 New England Power Company 1983 1150 4
Fort Calhoun Unit 2 Omaha Public Power District; Nebraska Public Power District Nebraska 1983 1150 4
Jamesport Unit 2 Long Island Lighting Company New York 1983 1150 4
Sears Island Central Maine Power Company Maine 1200 4


1.4.5.20 Framatome Technologi es, Incorporated (FTI)
B/B-UFSAR 1.4-19 TABLE 1.4-3 (Cont'd)
 
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Taiwan Unit 6 Taiwan Power Company Taiwan 1983 950 3
FTI is located in Lync hburg, Virginia and has been providing services to the electric power i ndustry for over four decades.
Alvin W. Vogtle Unit 2 Georgia Power Company Georgia 1984 1113 4
FTI engineering services include the necessary expertise, experience, and NRC-appr oved computer codes and methodologies to support the transient analysis of the Un it steam generators.
Marble Hill Unit 2 Public Service Company of Indiana, Inc.;
1.4.5.21 Stone & Webster Engineers and Construc tors, Inc. (S&W)
Northern Indiana Public Service Company Indiana 1984 1150 4
S&W is located in Bost on, Massachusetts and has been providing services to the electric power industry for over 100 years. S&W has provided balance-o f-plant design-engineeri ng support services in support of the po wer uprate of the Byron and Braidwood units.
Shearon Harris Unit 1 Carolina Power and Light Co.
B/B-UFSAR  
North Carolina 1984 900 3
 
Sterling SNUPPS - Rochester Gas and Electric Corporation; Central Hudson Gas and Electric Corporation; Niagara Mohawk Power Corporation; Orange and Rockland Utilities, Inc.
1.4-9 REVISION 8
New York 1984 1150 4
- DECEMBER 2000 TABLE 1.4-1 EXELON GENERATION COMPAN Y'S NUCLEAR POWER PLANTS IN SERVICE OR UNDER CONSTRUCTION UNIT NOMINAL GROSS 1 RATING (MWe)
Atlantic Unit 1 (O.P.S.)
SCHEDULED COMMERCIAL SERVICE DATE    Dresden 1 210 1960 Dresden 2 850 1972 Dresden 3 850 1972 Quad-Cities 1 850 1972 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 La Salle 1 1122 1978 La Salle 2 1122 1979 Byron 1 1175 1985 Byron 2 1175 1987 Braidwood 1 1175 1988 Braidwood 2 1175 1988 1Note that this is a gross rating, not a net rating.
Public Service Electric and Gas Company; Atlantic City Electric Co.;
B/B-UFSAR 1.4-10 REVISION 1 - DECEMBER 1989 TABLE 1.4-2 NUCLEAR POWER PLANTS COMPLETED OR CURRENTLY UNDER DESIGN BY SARGENT & LUNDY UNIT NOMINAL GROSS 2 RATING (MWe)
Jersey Central Power and Light Company New Jersey 1985 1150 4
YEAR OF POWER OPERATION    EBWR 5 1956 Elk River 22 1962 La Crosse 60 1967 SEFOR 20 (MWt) 1969 Dresden 2 850 1969 Dresden 3 850 1971 Quad-Cities 1 850 1971 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 Fort St. Vrain, Unit 1 330 1973 Enrico Fermi, Unit 2 1200 1988 La Salle County Station, Unit 1 1122 1979 La Salle County Station, Unit 2 1122 1980 Byron Station, Unit 1 1175 1985 Byron Station, Unit 2 1175 1987 Braidwood Statio n, Unit 1 1175 1988 Braidwood Statio n, Unit 2 1175 1988 Clinton Power Station, Unit 1 992 1981 Kaiseraugst 992 1982 2Note that this is a gross rating, not a net rating.
NEP-2 New England Power Company 1985 1150 4
B/B-UFSAR 1.4-11 TABLE 1.4-3 WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTS PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS        Shippingport Duquesne Light Company; Energy Research & Development Administration Pennsylvania 1957 90 4 Yankee-Rowe Yankee Atomic Electric Company Massachusetts 1961 175 4 Trio Vercellese (Enrico Fermi)
South Dade Unit 1 Florida Power and Light Co.
Ente Nazionale per L'Energia Elettrica (ENEL) Italy 1965 260 4        Chooz (Ardennes)
Florida 1985 1150 4
Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA) France 1967 305 4        San Onofre Unit 1 Southern California Edison Co.; San Diego Gas and Electric Co. California 1968 450 3 Haddam Neck (Connecticut Yankee)
Sundesert Unit 1 San Diego Gas and Electric Co.
Connecticut Yankee Atomic Power Company Connecticut 1968 575 4 Jose Cabrera-Zorita  Union Electrica, S.A. Spain 1969 153 1 Beznau Unit 1 Nordostschweizerische Krafwerke AG (NOK) Switzerland 1969 350 2 Robert Emmett Ginna  Rochester Gas and Electric Corporation New York 1970 490 2 Mihama Unit 1 The Kansai Electric Power Company, Inc. Japan 1970 320 2        Point Beach Unit 1 Wisconsin Electric Power Co.; Wisconsin Michigan Power Co. Wisconsin 1970 497 2 H. B. Robinson Unit 2  Carolina Power and Light Co. South Carolina 1971 707 3 B/B-UFSAR 1.4-12 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
California 1985 950 3  
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS       Beznau Unit 2 Nordostschweizerische Kraftwerke AG (NOK) Switzerland 1972 350 2 Point Beach Unit 2 Wisconsin Electric Power Co.; Wisconsin Michigan Power Co. Wisconsin 1972 497 2 Surry Unit 1 Virginia Electric and Power Co. Virginia 1972 822 3 Turkey Point Unit 3  Florida Power and Light Co. Florida 1972 745 3 Indian Point Unit 2  Consolidated Edison Company of New York, Inc. New York 1973 873 4 Prairie Island Unit 1  Northern States Power Company Minnesota 1973 530 2 Turkey Point Unit 4  Florida Power and Light Co. Florida 1973 745 3 Surry Unit 2 Virginia Electric and Power Co. Virginia 1973 822 3 Zion Unit 1 Exelon Generation Company Illinois 1973 1050 4 Kewaunee Wisconsin Public Service Corp.; Wisconsin Power and Light Co.;
Madison Gas and Electric Co. Wisconsin 1974 560 2 Prairie Island Unit 2  Northern States Power Company Minnesota 1974 530 2 Takahama Unit 1 The Kansai Electric Power Company, Inc. Japan 1974 781 3        Zion Unit 2 Exelon Generation Company Illinois 1974 1050 4
 
B/B-UFSAR 1.4-13 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Doel Unit 1 Indivision Doel Belgium 1975 390 2 Doel Unit 2 Indivision Doel Belgium 1975 390 2 Donald C. Cook Unit 1  Indiana and Michigan Electric Company (AEP) Michigan 1975 1060 4        Ringhals Unit 2 Statens Vattenfallsverk (SSPB) Sweden 1975 822 3 Almaraz Unit 1 Unit Electrica, S.A.; Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A. Spain 1976 902 3        Beaver Valley Unit 1  Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Company Pennsylvania 1976 852 3 Diablo Canyon Unit 1  Pacific Gas and Electric Co. California 1976 1084 4 Indian Point Unit 3  Consolidated Edison Company of New York, Inc. New York 1976 965 4 Lemoniz Unit 1 Iberduero, S.A. Spain 1976 902 3 Salem Unit 1 Public Service Electric and Gas Company; Exelon Generation Company; Atlantic City Electric Co.;
Delmarva Power and Light Co. New Jersey 1976 1090 4 B/B-UFSAR 1.4-14 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Trojan  Portland General Electric Co.; Eugene Water and Electric Board; Pacific Power and Light Company Oregon 1976 1130 4        Almaraz Unit 2 Union Electrica, S.A.; Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A. Spain 1977 902 3        Asco Unit 1 Fuerzas Electricas de Cataluna, S.A. (FESCA) Spain 1977 902 3        Diablo Canyon Unit 2  Pacific Gas and Electric Co. California 1977 1106 4 Joseph M. Farley Unit 1  Alabama Power Company Alabama 1977 829 3 Ko-Ri Unit 1 Korea Electric Power Co., Ltd. Korea 1977 564 2 North Anna Unit 1 Virginia Electric and Power Co. Virginia  1977 898 3 North Anna Unit 2 Virginia Electric and Power Co. Virginia 1977 898 3 Ohi Unit 1 The Kansai Electric Power Co., Inc. Japan 1977 1122 4 Ohi Unit 2 The Kansai Electric Power Co., Inc. Japan 1977 1122 4 Ringhals Unit 3 Statens Vattenfallsvert (SSPB) Sweden 1977 900 3 Sequoyah Unit 1 Tennessee Valley Authority Tennessee 1977 1148 4 Angra dos Reis Unit 1  Furnas-Centrais Electricas, S.A. Brazil 1978 626 2
 
B/B-UFSAR 1.4-15 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Asco Unit 2 Fuerzas Electricas de  Cataluna, S.A. (FESCA);
Empresa Nacional Hidroelectrica del Ribagorzana, S.A. (ENHER); Fuerzas Hidroelectricas del  Segre, S.A.;
Hidroelectrica de Cataluna, S.A. Spain 1978 902 3        Donald C. Cook Unit 2  Indiana and Michigan Electric Company (AEP) Michigan 1978 1060 4        Lemoniz Unit 2 Iberduero, S.A. Spain 1978 902 3 Sequoyah Unit  2 Tennessee Valley Authority Tennessee 1978 1148 4 Watts Bar Unit 1 Tennessee Valley Authority Tennessee 1978 1177 4 William B. McGuire Unit 1  Duke Power Company North Carolina 1978 1180 4 Joseph M. Farley Unit 2  Alabama Power Company Alabama 1979 829 3 Krsko  Savske Elektrarne, Ljubljana, Slovenia, Elektroprivreda, Zagreb, Croatia Yugoslavia 1979 615 2 Ringhals Unit 4 Statens Vattenfallsvert (SSPD) Sweden 1979 900 3
 
B/B-UFSAR 1.4-16 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Salem Unit 2 Public Service Electric and  Gas Company; Exelon Generation Company Atlantic City Electric Co.; Delmarva Power and Light Co. New Jersey 1979 1115 4 Virgil C. Summer South Carolina Electric and  Gas Company South Carolina 1979 900 3 Watts Bar Unit 2 Tennessee Valley Authority Tennessee 1979 1177 4 William B. McGuire Unit 2  Duke Power Company North Carolina 1979 1180 4 Byron Unit 1 Exelon Generation Company Illinois 1981 1120 4 Comanche Peak Unit 1  Texas Utilities Generating Co. Texas 1980 1150 4 Seabrook Unit 1 Public Service Company of    New Hampshire; United Illuminating Company New Hampshire 1980 1200 4 South Texas Project Unit 1  Houston Lighting and Power Co.;
Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Texas 1980 1250 4 B/B-UFSAR 1.4-17 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Beaver Valley Unit 2  Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Co.;
Cleveland Electric  Illuminating Company; Toledo Edison Company Pennsylvania 1981 852 3 Braidwood Unit 1 Exelon Generation Company Illinois 1981 1120 4 Callaway Unit 1 SNUPPS - Union Electric Co. Missouri 1981 1150 4 Catawba Unit 1 Duke Power Company South Carolina 1981 1153 4 Jamesport Unit 1 Long Island Lighting Company New York 1981 1150 4 Ko-Ri Unit 2 Korea Electric Power Co., Ltd. Korea 1981 605 2 NORCO  Puerto Rico Water Resources Authority Puerto Rico
- 583 2        Braidwood Unit 2 Exelon Generation Company Illinois 1982 1120 4 Byron Unit 2 Exelon Generation Company Illinois 1982 1120 4 Catawba Unit 2 Duke Power Company South Carolina 1982 1153 4 Comanche Peak Unit 2  Texas Utilities Generating Co. Texas 1982 1150 4 Marble Hill Unit 1  Public Service Company of  Indiana, Inc.;
Northern Indiana Public Service Company Indiana 1982 1150 4 B/B-UFSAR 1.4-18 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Millstone Unit 3 Northeast Nuclear Energy Co. Connecticut 1982 1156 4 Seabrook Unit 3 Public Service Company of  New Hampshire; United Illuminating Company New Hampshire 1982 1200 4 South Texas Project Unit 2  Houston Lighting and Power Co.; Central Power and Light Co.; City Public Service of  San Antonio; City of Austin, Texas Texas 1982 1250 4        Taiwan Unit 5 Taiwan Power Company Taiwan 1982 950 3 Wolf Creek Unit 1 SNUPPS - Kansas Gas and  Electric Company; Kansas City Power and Light Company Kansas 1982 1150 4        Alvin W. Vogtle Unit 1  Georgia Power Company Georgia 1983 1113 4 Callaway Unit 2 SNUPPS - Union Electric Company Missouri 1983 1150 4 NEP-1  New England Power Company
- 1983 1150 4 Fort Calhoun Unit 2  Omaha Public Power District; Nebraska Public Power District Nebraska 1983 1150 4        Jamesport Unit 2 Long Island Lighting Company New York 1983 1150 4 Sears Island Central Maine Power Company Maine    - 1200 4 B/B-UFSAR 1.4-19 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Taiwan Unit 6 Taiwan Power Company Taiwan 1983 950 3 Alvin W. Vogtle Unit 2 Georgia Power Company Georgia 1984 1113 4 Marble Hill Unit 2 Public Service Company   of Indiana, Inc.;
Northern Indiana Public   Service Company Indiana 1984 1150 4       Shearon Harris Unit 1 Carolina Power and Light Co. North Carolina 1984 900 3 Sterling SNUPPS - Rochester Gas and   Electric Corporation; Central Hudson Gas and Electric Corporation; Niagara Mohawk Power Corporation; Orange and Rockland Utilities, Inc. New York 1984 1150 4 Atlantic Unit 1 (O.P.S.)
Public Service Electric and   Gas Company; Atlantic City Electric Co.;
Jersey Central Power and   Light Company New Jersey 1985 1150 4 NEP-2 New England Power Company  
- 1985 1150 4 South Dade Unit 1 Florida Power and Light Co. Florida 1985 1150 4 Sundesert Unit 1 San Diego Gas and Electric Co. California 1985 950 3  


B/B-UFSAR 1.4-20 TABLE 1.4-3 (Cont'd)
B/B-UFSAR 1.4-20 TABLE 1.4-3 (Cont'd)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS       Tyrone Unit 1 SNUPPS - Northern States   Power Company Wisconsin 1985 1150 4 Shearon Harris Unit 2 Carolina Power and Light Co. North Carolina 1986 900 3 South Dade Unit 2 Florida Power and Light Co. Florida 1986 1150 4 Atlantic No. (O.P.S.) Public Service Electric and   Gas Company; Atlantic City Electric Co.;
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Tyrone Unit 1 SNUPPS - Northern States Power Company Wisconsin 1985 1150 4
Jersey Central Power and Light Company New Jersey 1987 1150 4 Shearon Harris Unit 4 Carolina Power and Light Co. North Carolina 1988 900 3 Sundesert Unit 2 San Diego Gas and Electric Co. California 1988 950 3 Sayago Unit 1 Iberduero, S.A. Spain 1980's 1000 3 Sayago Unit 4 Iberduero, S.A. Spain 1980's 1000 3 Shearon Harris Unit 3 Carolina Power and Light Co. North Carolina 1990 900 3 Unassigned Unit 1 (O.P.S.)
Shearon Harris Unit 2 Carolina Power and Light Co.
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1990 1150 4 Unassigned Unit 2 (O.P.S.)
North Carolina 1986 900 3
South Dade Unit 2 Florida Power and Light Co.
Florida 1986 1150 4
Atlantic No. (O.P.S.)
Public Service Electric and Gas Company; Atlantic City Electric Co.;
Jersey Central Power and Light Company New Jersey 1987 1150 4
Shearon Harris Unit 4 Carolina Power and Light Co.
North Carolina 1988 900 3
Sundesert Unit 2 San Diego Gas and Electric Co.
California 1988 950 3
Sayago Unit 1 Iberduero, S.A.
Spain 1980's 1000 3
Sayago Unit 4 Iberduero, S.A.
Spain 1980's 1000 3
Shearon Harris Unit 3 Carolina Power and Light Co.
North Carolina 1990 900 3
Unassigned Unit 1 (O.P.S.)
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1990 1150 4
Unassigned Unit 2 (O.P.S.)
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1992 1150 4  
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1992 1150 4  


B/B-UFSAR 1.5-1 1.5 REQUIREMENTS FOR FURTH ER TECHNICAL INFORMATION The design of the Byron/Braidwood units is bas ed upon proven concepts which have been develop ed and success fully applied to the design of pressurized water reactor system
B/B-UFSAR 1.5-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION The design of the Byron/Braidwood units is based upon proven concepts which have been developed and successfully applied to the design of pressurized water reactor systems. There are currently no areas of research and development which are required for operation of this plant.
: s. There are currently no areas of resear ch and developme nt which are required for operati on of this plant.
At the time of issuance of construction permits for the Byron/
At the time of issuance of construction permits for the Byron/
Braidwood units, the Prelimina ry Safety Analys is Report (PSAR) and the standard design report which it referenced, RESAR-3, identified certain res earch and development pr ograms which were incomplete. These p rograms, which have been successfully completed, have provid ed technical informati on which has been used either to demonst rate the safety of design, more sharply define margins of conser vatism, or lead to design improvements.
Braidwood units, the Preliminary Safety Analysis Report (PSAR) and the standard design report which it referenced, RESAR-3, identified certain research and development programs which were incomplete. These programs, which have been successfully completed, have provided technical information which has been used either to demonstrate the safety of design, more sharply define margins of conservatism, or lead to design improvements.
Reference 1 presents descriptions of those safety-related research and development programs which have been carried out for, by, or in conjunc tion with Westinghouse Nuclear Energy Systems, and which are applicable to Wes tinghouse pressurized water reactors. The discussion which follows documents the completion of the construction permit stage rese arch programs.
Reference 1 presents descriptions of those safety-related research and development programs which have been carried out for, by, or in conjunction with Westinghouse Nuclear Energy Systems, and which are applicable to Westinghouse pressurized water reactors. The discussion which follows documents the completion of the construction permit stage research programs.
1.5.1 Programs Required for Plant Operation Two programs were iden tified as required for plant design and operation in the PSAR:  
1.5.1 Programs Required for Plant Operation Two programs were identified as required for plant design and operation in the PSAR:
: a. core stability evaluation and  
: a. core stability evaluation and
: b. fuel rod burst program.  
: b. fuel rod burst program.
 
Both programs are complete. The fuel rod burst program was completed at the time of the PSAR. The core stability evaluation program was not. A discussion of the core stability evaluation program follows.
Both programs are co mplete. The fuel rod burst program was completed at the time of the PSA R. The core sta bility evaluation program was not. A disc ussion of the core s tability evaluation program follows.
1.5.1.1 Core Stability Evaluation The program to establish means for the detection and control of potential xenon oscillations and for the shaping of the axial power distribution for improved core performance has been satisfactorily completed. See item 1, Reference 2, for a further discussion of the tests and results.
1.5.1.1 Core Sta bility Evaluation
1.5.2 Other Programs Not Required for Plant Operation The following programs were not complete at the time of the PSAR but are now satisfactorily complete.  
 
The program to establish means for the detection and control of potential xenon oscill ations and for the shaping of the axial power distribution for improved core per formance has been satisfactorily completed.
See item 1, Referen ce 2, for a further discussion of the tests and results.
1.5.2 Other Programs N ot Required for Plant Operation
 
The following programs w ere not complete at the time of the PSAR but are now satisfac torily complete.
 
B/B-UFSAR 1.5-2 1.5.2.1 Fuel Development Program for Operation at High Power Densities The program to demonst rate the satisfact ory operation of fuel at high burnup and p ower densities has been satisfactorily completed. See item 8, Reference 2, for a further discussion of the program a nd its results.
 
1.5.2.2 Blowdown Forces Program Westinghouse has completed B LODWN-2, an improved digital computer program for the calcula tion of local fluid pressures, flows and density transients in the primary coolant systems during a LOCA.  


BLODWN-2 is used to evaluate the effects of blowdown forces in this application. Refer to it em 15 in Reference 4 for a further discussion of the tests and results.
B/B-UFSAR 1.5-2 1.5.2.1 Fuel Development Program for Operation at High Power Densities The program to demonstrate the satisfactory operation of fuel at high burnup and power densities has been satisfactorily completed. See item 8, Reference 2, for a further discussion of the program and its results.
1.5.2.3 Blowdown Heat Transfer Testing (Form erly Titled Delayed Departure From N ucleate Boiling)
1.5.2.2 Blowdown Forces Program Westinghouse has completed BLODWN-2, an improved digital computer program for the calculation of local fluid pressures, flows and density transients in the primary coolant systems during a LOCA.
The NRC Acceptance Cri teria for Emergency Core Cooling Systems for Light-Water Powe r Reactors was issued in Section 50.46 of 10 CFR 50 on December 28, 19
BLODWN-2 is used to evaluate the effects of blowdown forces in this application. Refer to item 15 in Reference 4 for a further discussion of the tests and results.
: 73. It defines the basis and conservative assumptio ns to be used in the evaluation of the performance of emergency core cooling systems (ECCS).
1.5.2.3 Blowdown Heat Transfer Testing (Formerly Titled Delayed Departure From Nucleate Boiling)
Westinghouse believes that some of the conservatism of the criteria is associated w ith the manner in wh ich transient DNB phenomena are treated in the e valuation models.
The NRC Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Power Reactors was issued in Section 50.46 of 10 CFR 50 on December 28, 1973. It defines the basis and conservative assumptions to be used in the evaluation of the performance of emergency core cooling systems (ECCS).
Transient critical heat flux data presented at the 1972 specialists meeting of the Committee on Reactor Safe ty Technology (C REST) indicated that the time to DNB can be dela yed under transi ent conditions.
Westinghouse believes that some of the conservatism of the criteria is associated with the manner in which transient DNB phenomena are treated in the evaluation models. Transient critical heat flux data presented at the 1972 specialists meeting of the Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be delayed under transient conditions.
To demonstrate the c onservatism of the E CCS evaluation models, Westinghouse initiated a program to experiment ally simulate the blowdown phase of a LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blow down Heat Transfer Program, which was start ed early in 1976. T esting was completed in 1979. A DNB corr elation developed by Wes tinghouse from these test results is used in the ECCS analyses for Byron/Braidwood.
To demonstrate the conservatism of the ECCS evaluation models, Westinghouse initiated a program to experimentally simulate the blowdown phase of a LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blowdown Heat Transfer Program, which was started early in 1976. Testing was completed in 1979. A DNB correlation developed by Westinghouse from these test results is used in the ECCS analyses for Byron/Braidwood.
Objective The objective of the blo wdown heat transfer te st was to determine the time that DNB occurs under LOCA conditions. Thi s information was used to confirm a new Westinghou se transient DNB correlation.
Objective The objective of the blowdown heat transfer test was to determine the time that DNB occurs under LOCA conditions. This information was used to confirm a new Westinghouse transient DNB correlation.
The steady-state DNB data obtain ed from 15x15 and 17x17 test programs was used to assure th at the geometrical differences between the two fuel a rrays is correctly treat ed in the transient correlations.  
The steady-state DNB data obtained from 15x15 and 17x17 test programs was used to assure that the geometrical differences between the two fuel arrays is correctly treated in the transient correlations.  


B/B-UFSAR 1.5-3 Program The program was divided into two phases.
B/B-UFSAR 1.5-3 Program The program was divided into two phases. The Phase I tests started from steady-state conditions, with sufficient power to maintain nucleate boiling throughout the bundle, and progressed through controlled ramps of decreasing test section pressure or flow initiated DNB. By applying a series of controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pressures relevant to a PWR blowdown.
The Phase I tests started from steady-st ate conditions, with sufficient power to maintain nucleate boiling throughout the bundle, and progressed  
Phase I provided separate-effects data for heat transfer correlation development.
Typical parameters used for Phase I testing are shown in Table 1.5-1.
Phase II simulated PWR behavior during a LOCA to permit definition of the time delay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after establishment of typical steady-state operating conditions. The fluid transient was then initiated, and the rod power decay was programmed in such a manner as to simulate the actual heat input of fuel rods. The test was terminated when the heater rod temperatures reached a predetermined limit.
Typical parameters used for Phase II testing are shown in Table 1.5-2.
Test Description The experimental program was conducted in the J-Loop at the Westinghouse Forest Hills Facility with a full length 5x5 rod bundle simulating a section of a 15x15 fuel assembly to determine DNB occurrence under LOCA conditions.
The heater rod bundles used in this program were internally-heated rods, capable of a maximum linear power of 18.8 kW/ft, with a total power of 135 kW (for extended periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribution. Each rod was adequately instrumented with a total of 12 clad thermocouples.
Results The experiments in the DNB facility resulted in cladding temperature and fluid properties measured as a function of time throughout the blowdown range from 0 to 20 seconds.
Facility modifications and installation of the initial test bundle were completed. A series of shakedown tests in the


through controlled ram ps of decreasing test se ction pressure or flow initiated DNB.
B/B-UFSAR 1.5-4 REVISION 1 - DECEMBER 1989 J-Loop were performed. These tests provided data for instrumentation calibration and check-out, and provided information regarding facility control and performance. Initial program tests were performed during the first half of 1975.
By applying a series of controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pres sures relevant to a PWR blowdown.  
Under the sponsorship of EPRI, testing was reinitiated during 1976 on the same test bundle. The testing was terminated in November 1976 and plans were made for a new test bundle and further testing during 1978-1979. These tests were completed in December of 1979.
1.5.3 References
: 1.
F. T. Eggleston, "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, October 1978.
: 2.
F. T. Eggleston, "Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8768.
Spring 1976 Edition.
: 3.
"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8458. Fall 1977 Edition.
: 4.
"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8004. Fall 1972 Edition.  


Phase I provided separate-effects data for h eat transfer correlation development.  
B/B-UFSAR 1.5-5 TABLE 1.5-1 BLOWDOWN HEAT TRANSFER PHASE I TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 1250 to 2250 psia Test section mass velocity 1.12 to 2.5x106 lb/hr-ft2 Core inlet temperature 550° F to 600° F Maximum heat flux 306,000 to 531,000 Btu/hr-ft2 TRANSIENT RAMP CONDITIONS Pressure decrease 0 to 350 psia/sec and subcooled depressurization from 2250 psia Flow decrease 0 to 100%/sec Inlet enthalpy constant


Typical parameters used for Phase I test ing are shown in Table 1.5-1.
B/B-UFSAR 1.5-6 TABLE 1.5-2 BLOWDOWN HEAT TRANSFER PHASE II TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 2250 psia Test section mass velocity 2.5x106 lb/hr-ft2 Inlet coolant temperature 545° F Maximum heat flux 531,000 Btu/hr-ft2 TRANSIENT CONDITIONS Simulated break Double-ended cold leg guillotine breaks
Phase II simulated P WR behavior during a LOCA to permit definition of the time d elay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after establishment of typical steady-stat e operating cond itions. The fluid transient was th en initiated, and the rod power decay was programmed in such a m anner as to simulate t he actual heat input of fuel rods. The test was terminated when the heater rod temperatures reached a predetermined limit.


Typical parameters u sed for Phase II testing are shown in Table 1.5-2.  
B/B-UFSAR 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCES Table 1.6-1 lists topical reports which provide information additional to that provided in this UFSAR and which have been filed separately with the Nuclear Regulatory Commission (NRC) in support of this and similar applications.
A legend to the review status code letters follows:
A
- NRC review complete; NRC acceptance letter issued.
AE
- NRC accepted as part of the Westinghouse Emergency Core Cooling System (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.
B
- Submitted to the NRC as background information; not undergoing formal NRC review.
O
- On file with NRC; older generation report with current validity; not actively under formal NRC review.
U
- Actively under formal NRC review.  


Test Description
B/B-UFSAR 1.6-2 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE REPORT REFERENCE SECTION(S)
REVIEW STATUS "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," WCAP-2048, July 1962 4.3 0
"Single Phase Local Boiling and Bulk Boiling Pressure Drop Correlations," WCAP-2850 (Proprietary), April 1966 and WCAP-7916 (Non-Proprietary), June 1972 4.4 0
"In-Pile Measurement of UO2 Thermal Conductivity," WCAP-2923, 1966 4.4 0
"Hydraulic Tests of the San Onofre Reactor Model," WCAP-3269-8, June 1964 4.4 0
"LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM - 7094,"
WCAP-3269-26, September 1963 4.3, 4.4 15.0, 15.4 0
"Saxton Core II Fuel Performance Evaluation,"
WCAP-3385-56, Part II, "Evaluation of Mass Spectrometric and Radiochemical Analyses of Irradiated Saxton Plutonium Fuel," July 1970 4.3, 4.4 0
"Xenon-Induced Spatial Instabilities in Large PWRs," WCAP-3680-20, (EURAEC-1974)
March 1968 4.3 0
"Control Procedures for Xenon-Induced X-Y Instabilities in Large PWR's,"
WCAP-3680-21, (EURAEC-2111) February 1969 4.3 0
"Xenon-Induced Spatial Instabilities in Three-Dimensions," WCAP-3680-22, (EURAEC-2116) September 1969 4.3 0
"Pressurized Water Reactor pH - Reactivity Effect Final Report," WCAP-3698-8, (EURAEC-2074) October 1968 4.3 0
"PUO2 - UO2 Fueled Critical Experiments,"
WCAP-3726-I, July 1967 4.3 0


The experimental program was con ducted in the J-Loop at the Westinghouse Forest Hills Facili ty with a full length 5x5 rod bundle simulating a se ction of a 15x15 fuel as sembly to determine DNB occurrence under LOCA conditions.  
B/B-UFSAR 1.6-3 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Melting Point of Irradiated UO2,"
WCAP-6065, February 1965 4.2, 4.4 0
"Burnup Physics of Heterogeneous Reactor Lattices," WCAP-6069, June 1965 4.4 0
"LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS,"
WCAP-6073, April 1966 4.3 0
"Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium," WCAP-6086, August 1969 4.3 0
"Subchannel Thermal Analysis of Rod Bundle Cores," WCAP-7015, Revision 1, January 1969 4.4 0
"The PANDA Code," WCAP-7048 (Proprietary) and WCAP-7757 (Non-Proprietary), January 1975 4.3 A
"Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198-L (Proprietary), April 1969 and WCAP-7825 (Non-Proprietary), December 1971 4.3 0
"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7208 (Proprietary), September 1968 and WCAP-7811, (Non-Proprietary), December 1971 4.3 "The TURTLE 24.0 Diffusion Depletion Code,"
WCAP-7213 (Proprietary) and WCAP-7758 (Non-Proprietary), January 1975 4.3, 15.0 15.4 A
"Core Power Capability in Westinghouse PWRs,"
WCAP-7267-L (Proprietary), October 1969 and WCAP-7809 (Non-Proprietary), December 1971 4.3 "Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors,"
WCAP-7306, April 1969 15.4 "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7308, December 1971 4.3 A


The heater rod bundl es used in this prog ram were internally-heated rods, cap able of a maximum line ar power of 18.8 kW/ft, with a total power of 135 kW (for exte nded periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribut ion. Each rod was adequately instrumented with a total of 12 clad thermocouples.
B/B-UFSAR 1.6-4 TABLE 1.6-1 (Cont'd)
Results The experiments in t he DNB facility re sulted in cladding temperature and fluid pr operties measured as a function of time throughout the blowd own range from 0 to 20 seconds.
REPORT REFERENCE SECTION(S)
Facility modifications and installation of the initial test bundle were completed.
REVIEW STATUS "Application of the THINC Program to PWR Design," WCAP-7359-L (Proprietary), August 1969 and WCAP-7838 (Non-Proprietary),
A series of shak edown tests in the
January 1972 4.4 O
"Seismic Testing of Electrical and Control Equipment," WCAP-7397-L (Proprietary) and WCAP-7817 (Non-Proprietary), December 1971 3.10 O
"Seismic Testing of Electrical and Control Equipment (WCID Process Control Equipment),"
WCAP-7397-L, Supplement 1 (Proprietary) and WCAP-7817, Supplement 1 (Non-Proprietary),
December 1971 3.10 O
"Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," WCAP-7477-L (Proprietary), March 1970 and WCAP-7735 (Non-Proprietary), August 1971 5.2 A
"Radiological Consequences of a Fuel Handling Accident," WCAP-7518-L (Proprietary) and WCAP-7828 (Non-Proprietary), June 1970 15.7 O
"Seismic Vibration Testing with Sine Beats,"
WCAP-7558, October 1972 3.10 O
"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975 15.4 A
"Dynamic Fracture Toughness Properties of Heavy Section A533 Grade B Class 1 Steel Plate," WCAP-7623, December 1970 5.4 O
"Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L (Proprietary) and WCAP-7755 (Non-Proprietary), January 1975 4.4 A
"DNB Tests Results for New Mixing Vane Grids (R)," WCAP-7695-L (Proprietary) and WCAP-7958 (Non-Proprietary) and Addendum, January 1975 4.4 A


B/B-UFSAR 1.5-4 REVISION 1 - DECEMBER 1989 J-Loop were performed. These tests provided data for instrumentation calibr ation and check-ou t, and provided information regarding fa cility control a nd performance. Initial program tests were performed during the first half of 1975.
B/B-UFSAR 1.6-5 TABLE 1.6-1 (Cont'd)
Under the sponsorship of EPRI, t esting was rei nitiated during 1976 on the same test bu ndle. The testing was terminated in November 1976 and plans were m ade for a new test bundle and further testing during 1 978-1979. These tes ts were completed in December of 1979.  
REPORT REFERENCE SECTION(S)
REVIEW STATUS "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients,"
WCAP-7706, February 1973 4.6, 7.1 O
"Electric Hydrogen Recombiner for PWR Containments," WCAP-7709-L, Supplements 1 through 7 (Proprietary) and WCAP-7820, Supplements 1 through 7 (Non-Proprietary),
1971 through 1977 3.11, 6.2 A
"A Comprehensive Space-Time Dependent Analysis of Loss of Coolant (SATAN-IV Digital Code),"
WCAP-7750, August 1971 3.6 O
"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, October 1971 15.2 O
"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June 1972 5.2 O
"Behavior of Austenitic Stainless Steel in Post Hypothetical Loss of Coolant Accident Environment," WCAP-7798-L (Proprietary) and WCAP-7803 (Non-Proprietary), January 1972 6.1 O
"Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800, Revision 4-A, April 1975 4.2, 17 A
"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"
WCAP-7806, December 1971 4.3 B
"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7811, December 1971 4.3 O
"Seismic Testing of Electrical and Control Equipment (Low Seismic Plants)," WCAP-7817, Supplements 1-8, December 1971-March 1974 3.10 O


====1.5.3 References====
B/B-UFSAR 1.6-6 TABLE 1.6-1 (Cont'd)
: 1. F. T. Eggleston, "Safety-Rel ated Research and Development for Westinghouse Pressurized W ater Reactors, Program  
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Evaluation of Steam Generator Tube, Tubesheet and Divider Plate Under Combined LOCA Plus SSE Conditions," WCAP-7832, December 1973 5.4 A
"Inlet Orificing of Open PWR Cores,"
WCAP-7836, January 1972 4.4 B
"Neutron Shielding Pads," WCAP-7870, May 1972 3.9 A
"LOFTRAN Code Description," WCAP-7907, June 1972 5.2, 15.0 15.1, 15.2, 15.3, 15.4, 15.5, 15.6 A
"FACTRAN - A FORTRAN-IV Code for Thermal Transients in a UO2 Fuel Rod," WCAP-7908, June 1972 15.0, 15.2 15.3, 15.4 A
"MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop PWR System,"
WCAP-7909, June 1972 6.3 O
"Power Peaking Factors," WCAP-7912-L (Proprietary) and WCAP-7912 (Non-Proprietary), January 1975 and Supplement 4.3, 4.4 A
"Damping Values of Nuclear Power Plant Components," WCAP-7921, May 1974 lA, 3.7 A
"Basis for Heatup and Cooldown Limit Curves," WCAP-7924, April 1975 5.3 A
"Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid,"
WCAP-7941-L (Proprietary) and WCAP-7959 (Non-Proprietary), January 1975 4.4 A
"Fuel Assembly Safety Analysis for Combined Seismic and Loss of Coolant Accident, 15x15,"
WCAP-7950, July 1972 3.7 A
"THINC-IV An Improved Program for Thermal and Hydraulic Analysis of Rod Bundle Cores,"
WCAP-7956, June 1973 4.4 A


Summaries," WCAP-876 8, October 1978.  
B/B-UFSAR 1.6-7 REVISION 9 - DECEMBER 2002 TABLE 1.6-1 (Cont'd)
: 2. F. T. Eggleston, "Safety-Relat ed Research and Development for Westinghouse PWRs Program Summaries," WCAP-8768.
REPORT REFERENCE SECTION(S)
Spring 1976 Edition.
REVIEW STATUS "Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor," WCAP-7964, June 1971 4.3 O
: 3. "Safety-Related Rese arch and Development for Westinghouse PWRs Program Summaries," WCA P-8458. Fall 1977 Edition.  
"TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979 (Proprietary) and WCAP-8028 (Non-Proprietary), January 1975 15.0, 15.4 A
: 4. "Safety-Related Rese arch and Development for Westinghouse PWRs Program Summaries," WCA P-8004. Fall 1972 Edition.  
"WIT-6 Reactor Transient Analysis Computer Program Description," WCAP-7980, November 1972 15.0, 15.4 A
"Application of Modified Spacer Factor to "L" Grid Typical and Cold Wall Cell DNB,"
WCAP-7988 (Proprietary) and WCAP-8030 (Non-Proprietary), October 1972 4.4 A
"Application of the THINC-IV Program to PWR Design," WCAP-8054 (Proprietary) and WCAP-8195 (Non-Proprietary), October 1973 4.4 A
"Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop,"
WCAP-8082 (Proprietary) and WCAP-8172 (Non-Proprietary), January 1975 3.6 A
"Reactor Coolant Pump Integrity in LOCA,"
WCAP-8163, September 1973 lA, 5.4 O
"Calculational Model for Core Reflooding After a Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974 15.6 A
"Effect of Local Heat Flux Spikes on DNB in Non-Uniform Heated Rod Bundles," WCAP-8174 (Proprietary) and WCAP-8202, (Non-Proprietary), August 1973 4.4 A
"WFLASH, A FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWR," WCAP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision 1 (Non-Proprietary), July 1974 15.6 A


B/B-UFSAR 1.5-5 TABLE 1.5-1 BLOWDOWN HEAT TRANSFER P HASE I TEST PARAMETERS
B/B-UFSAR 1.6-8 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Fuel Densification Experimental Results and Model for Reactor Application,"
WCAP-8218 (Proprietary) and WCAP-8219 (Non-Proprietary), March 1975 4.1, 4.2, 4.3, 4.4 A
"Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236 (Proprietary), December 1973 and WCAP-8288 (Non-Proprietary), January 1974 and Addenda 3.7, 4.2 A
"Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March 1974 and WCAP-8288, Addendum 1 (Non-Proprietary), April 1974 3.7 A
"Documentation of Selected Westinghouse Structural Analysis Computer Codes,"
WCAP-8252, Revision 1, July 1977 3.6, 3.9 O
"Hydraulic Flow Test of the 17x17 Fuel Assembly," WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary), February 1974 4.2, 4.4 O
"Effect of 17x17 Fuel Assembly Geometry on DNB," WCAP-8296 (Westinghouse Proprietary) and WCAP-8927 (Non-Proprietary), February 1975 4.4 A
"The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing," WCAP-8298 (Proprietary) and WCAP-8299 (Non-Proprietary),
January 1975 4.4 A
"LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974 15.0, 15.6 AE SATAN-IV Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant,"
WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974 15.0, 15.6 AE


PARAMETERS NOMINAL VALUE  INITIAL STEADY-S TATE CONDITIONS Pressure 1250 to 2250 psia
B/B-UFSAR 1.6-9 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests,"
WCAP-8303 (Proprietary) and WCAP-8317 (Non-Proprietary), July 1975 3.9 A
"Control of Delta Ferrite in Austenitic Stainless Steel Weldments," WCAP-8324-A, June 1974 lA, 5.2 A
"Containment Pressure Analysis Code (COCO),"
WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974 15.6 AE "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974 4.3, 4.6, 15.1, 15.2, 15.4, 15.8 O
"Westinghouse ECCS Evaluation Model -
Summary," WCAP-8339, July 1974 6.2, 15.6 AE "Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974 15.6 AE "Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP -8341 (Proprietary) and WCAP-8342 (Non-Proprietary), July 1974 lA(N), 17 A
"Effects of Fuel Densification Power Spikes on Clad Thermal Transients," WCAP-8359, July 1974 4.3 AE "Westinghouse Nuclear Energy Systems Division Quality Assurance Plan," WCAP-8370, Revision 9A, September 1977 1A, 17 A
"Qualification of Westinghouse Seismic Testing Procedure for Electrical Equipment Tested Prior to May 1974," WCAP-8373, August 1974 3.10 O
"Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-Proprietary),
July 1974 4.2 A


Test section mass velocity 1.12 to 2.5x10 6 lb/hr-ft 2
B/B-UFSAR 1.6-10 TABLE 1.6-1 (Cont'd)
Core inlet temperature 550° F to 600
REPORT REFERENCE SECTION(S)
° F Maximum heat flux 306,000 to 531,000 Btu/hr-ft 2 
REVIEW STATUS "Power Distribution Control and Load Following Procedures," WCAP-8385 (Proprietary) and WCAP-8403 (Non-Proprietary), September 1974 4.3, 4.4 A
"An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," WCAP-8424, Revision 1, June 1975 15.3 O
"17x17 Drive Line Components Tests - Phase IB, II, III, D-Loop Drop and Deflection,"
WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December 1974 3.9, 15.0 A
"Analysis of Data from the Zion (Unit 1)
THINC Verification Test," WCAP-8453-A (Proprietary), May 1976 and WCAP-8454 (Non-Proprietary), January 1975 4.4 A
"Westinghouse ECCS Evaluation Model -
Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-Proprietary), April 1974 15.6 AE "Incore Power Distribution Determination in Westinghouse Pressurized Water Reactors,"
WCAP-8498, July 1975 4.3 O
"UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations," WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary), April 1975 3.9 A
"Critical Heat Flux Testing of 17x17 Fuel Assembly Geometry with 22 Inch Spacing,"
WCAP-8536 (Proprietary) and WCAP-8537 (Non-Proprietary), May 1975 4.4 A
"Westinghouse ECCS - Four Loop Plant (17x17)
Sensitivity Studies," WCAP-8565 (Proprietary) and WCAP-8566 (Non-Proprietary), July 1975 15.6 A


TRANSIENT RAMP CONDITIONS
B/B-UFSAR 1.6-11 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Improved Thermal Design Procedure,"
WCAP-8567-P, July 1975 (Proprietary) and WCAP-8568, July 1975 (Non-Proprietary) 4.4, 15.0 A
"Augmented Startup and Cycle 1 Physics Program Supplement 1," WCAP-8575, June 1976 (Proprietary) and WCAP-8576, June 1976 (Non-Proprietary) and Supplements.
4.3 O
"The Application of Preheat Temperatures After Welding Pressure Vessel Steels,"
WCAP-8577, February 1976 lA A
"Failure Mode and Effects Analysis (FMEA) of the Engineered Safeguard Features Actuation System," WCAP-8584 (Proprietary) and WCAP-8760 (Non-Proprietary), April 1976 4.6 O
"Environmental Qualification of Westinghouse NSSS Class lE Equipment," WCAP-8587, September 1975 lA, 3.1O, 3.11 A
"Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary), November 1975 15.6 A
"Experimental Verification of Wet Fuel Storage Criticality Analyses," WCAP-8682 (Proprietary) and WCAP-8683 (Non-Proprietary), December 1975 4.3 B
"Fuel Rod Bowing," WCAP-8691 (Proprietary) and WCAP-8692 (Non-Proprietary),
December 1975 4.2 O
"Delta Ferrite in Production Austenitic Stainless Steel Weldments," WCAP-8693, January 1976 lA, 5.2 B
"MULTIFLEX - A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708 (Proprietary) and WCAP-8709 (Non-Proprietary), February 1976 3.9 A


Pressure decrease 0 to 350 psia/sec and subcooled depressurization
B/B-UFSAR 1.6-12 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS Foster, J. P., et al., Westinghouse Improved Performance Analysis and Design Model (PAD 4.0), WCAP-15063-P-A, Revision 1 with Errata, July 2000.
4.2 A
"New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, July 1976 (Proprietary) and WCAP-8763, July 1976 (Non-Proprietary) 4.4 A
"Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, Revision 2, October 1978 1.5, 4.2, 4.3 B
"Verification of Neutron Pad and 17x17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant," WCAP-8780, May 1976 3.9 B
"Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations,"
WCAP-8785, October 1976 4.2 "Hybrid B4C Absorber Control Rod Evaluation Report," WCAP-8846, October 1977 4.2, 15.0 15.3 A
"Westinghouse ECCS - Four Loop Plant (17x17)
Sensitivity Studies with Upper Head Fluid Temperature at Thot," WCAP-8865, May 1977 15.6 A
"7300 Series Process Control System Noise Tests," WCAP-8892-A, April 1977 7.1 A
"Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8963 (Proprietary), November 1976 and WCAP-8964 (Non-Proprietary), August 1977 4.2 A
"Westinghouse Emergency Core Cooling System Small Break October 1975 Model," WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary),
April 1977 15.6 A


from 2250 psia
B/B-UFSAR 1.6-13 REVISION 1 - DECEMBER 1989 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Failure Mode and Effects Analysis of the Solid State Full Length Rod Control System,"
WCAP-8976, September 1977 4.6 O
"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"
WCAP-9000-L, Revision 1 (Proprietary), July 1969 and WCAP-7806 (Non-Proprietary), December 1971.
4.3 "Axial Power Distribution Monitoring Using Four-Section Ex-Core Detectors," WCAP-9105 (Proprietary) and WCAP-9106 (Non-Proprietary),
July 1977 4.3 A
"Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCAs During Operation with One Loop Out of Service for Plants Without Loop Isolation Valves,"
WCAP-9166 (Proprietary) and WCAP-9167 (Non-Proprietary), February 1978 15.6 O
"Westinghouse Emergency Core Cooling System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9150 (Non-Proprietary), September 1977 15.6 O
"Properties of Fuel and Core Component Materials," WCAP-9179 (Proprietary), September 1977 and WCAP-9224 (Non-Proprietary) 4.2 O
"Westinghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220 (Proprietary Version), WCAP-9221 (Non-Proprietary Version), February 1978 15.6 A
"Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly," WCAP-9401 (Proprietary) and WCAP-9402 (Non-Proprietary), March 1979 4.1, 4.2, 4.4 A
"PALADON - Westinghouse Nodal Computer Code,"
WCAP-9485 (Proprietary) and WCAP-9486 (Non-Proprietary) December 1978 4.3 A


Flow decrease 0 to 100%/sec
B/B-UFSAR 1.6-14 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Reference Core Report 17x17 Optimized Fuel Assembly," WCAP-9500 (Non Proprietary),
July 1979 4, 15 A
"RELAP5/MOD2-B&W - An Advanced Computer Code for Light Water Reactor LOCA and non-LOCA Transient Analysis" BAW-10164, Revision 3 (non-proprietary), October 1996 15 A
"CONTEMPT - Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident,", BAW-10095A, Revision 1, April 1978 6
O Beacon Core Monitoring and Operations Support System, WCAP-12472 (Proprietary Class 2),
August 1994 4.3, 4.4, 7.7 A
Relaxation of Constant Axial Offset Control, FQ Surveillance Technical Specification, WCAP-10216-P-A, Revision 1A (Proprietary Class 2), February 1994 4.3, 4.4 A
VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, WCAP-14565-P-A (Proprietary) / WCAP-15306-NP-A (Non-Proprietary), October 1999 4.4, 15.0 A
Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications, WCAP-14565-P-A Addendum 2-P-A (Proprietary) / WCAP-15306-NP-A Addendum 2-NP-A (Non-Proprietary), April 2008 4.4 A
SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, WCAP-10698-P-A (Proprietary) and WCAP-10750-A (Non-Proprietary), August 1987 15.0 A


Inlet enthalpy constant
B/B-UFSAR 1.7-1 REVISION 9 - DECEMBER 2002 1.7 DRAWINGS The drawings cited in each UFSAR Chapter 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. References on the figures contained in the UFSAR to ComEd, CECo, and Commonwealth Edison will be revised to reflect the change in facility ownership to Exelon Generation Company when other changes to that figure are needed.
 
1.7.1 Electrical, Instrumentation, and Control Drawings Subsection 1.7.1 of the FSAR provides a list of electrical, instrumentation, and control drawings that were provided to the NRC during the initial licensing phase.
B/B-UFSAR 1.5-6 TABLE 1.5-2 BLOWDOWN HEAT TRANSF ER PHASE II TEST PARAMETERS
1.7.2 Drawings for Independent Structural Review Subsection 1.7.2 of the FSAR provides a list of the structural, architectural, mechanical loading and electrical loading drawings that were provided to the NRC to enable them to perform the Project Structural Review and the Independent Structural Review during the licensing phase.  
 
PARAMETERS NOMINAL VALUE  INITIAL STEADY-S TATE CONDITIONS Pressure 2250 psia
 
Test section mass velocity 2.5x10 6 lb/hr-ft 2
Inlet coolant temperature 545° F Maximum heat flux 531,000 Btu/hr-ft 2 
 
TRANSIENT CONDITIONS Simulated break
 
Double-ended cold leg
 
guillotine breaks
 
B/B-UFSAR 1.6-1 1.6 MATERIAL INCORPO RATED BY REFERENCES Table 1.6-1 lists topical repo rts which prov ide information additional to that p rovided in this UFSAR and which have been filed separately with the Nuclear Regulatory Com mission (NRC) in support of this and si milar applications.
 
A legend to the revi ew status code l etters follows:
A - NRC review complete; NRC acceptance letter issued.
AE - NRC accepted as part of the Westinghouse Emergency Core Cooli ng System (ECCS) evaluation model only; does not constitute acceptance for any purpo se other than for ECCS analyses. B - Submitted to t he NRC as background information; not undergo ing formal NRC review. O - On file with NRC; ol der generation report with current validity; not actively under formal
 
NRC review. U - Actively under f ormal NRC review.
 
B/B-UFSAR 1.7-1 REVISION 9 - DECEMBER 2002  
 
===1.7 DRAWINGS===
The drawings cited in each UFSAR Chapter are included as "General References" only; i.
e., refer to the d rawings to obtain additional detail or to obtain background information. These drawings are not part of the UFS AR. They are controlled by the Controlled Documents Program. References on the figures contained in the UFSAR to ComE d, CECo, and Com monwealth Edison will be revised to r eflect the change in facility ownership to Exelon Generation Company when other c hanges to that figure are needed. 1.7.1 Electrical, Instrumentat ion, and Control Drawings
 
Subsection 1.7.1 of the FSAR provides a list of electrical, instrumentation, and control dra wings that were provided to the NRC during the initi al licensing phase.
1.7.2 Drawings for Indepen dent Structural Review
 
Subsection 1.7.2 of the FSAR provides a list of the structural, architectural, mechanical loading and electrical loading drawings that were provided to the NR C to enable them to perform the Project Structural Rev iew and the Indepe ndent Structural Review during the lic ensing phase.  


B/B-UFSAR 1.7-2 REVISION 9 - DECEMBER 2002 TABLE 1.7-1 This Table has been intentionally deleted.  
B/B-UFSAR 1.7-2 REVISION 9 - DECEMBER 2002 TABLE 1.7-1 This Table has been intentionally deleted.  


B/B-UFSAR REVISION 9 - DECEMBER 2002  
B/B-UFSAR REVISION 9 - DECEMBER 2002 Pages 1.7-3 through 1.7-17 have been intentionally deleted.  
 
Pages 1.7-3 through 1.7-17 have been int entionally deleted.  
 
B/B-UFSAR  REVISION 9 - DECEMBER 2002
 
Figures 1.1-1 through 1.1-3 have been de leted intentionally.
 
B/B-UFSAR  REVISION 9 - DECEMBER 2002


Figures 1.2-1 through 1.
B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.1-1 through 1.1-3 have been deleted intentionally.  
2-17 have been d eleted intentionally.


B/B-UFSAR 1.0-i  REVISION 5 - DECEMBER 1994 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS
B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.2-1 through 1.2-17 have been deleted intentionally.


PAGE
B/B-UFSAR 1.0-i REVISION 5 - DECEMBER 1994 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS PAGE  


==1.0 INTRODUCTION==
==1.0 INTRODUCTION==
AND GENERAL DESCRIPTION OF PLANT 1.1-1
AND GENERAL DESCRIPTION OF PLANT 1.1-1  


==1.1 INTRODUCTION==
==1.1 INTRODUCTION==
1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Site and Environment 1.2-1 1.2.2 Nuclear Steam Supply System 1.2-1 1.2.3 Engineered Safety Features 1.2-2 1.2.4 Emergency Core Cooling System 1.2-3 1.2.5 Control and Instrumentation 1.2-3 1.2.6 Electrical System 1.2-4 1.2.7 Turbine and Auxiliaries 1.2-4 1.2.8 Fuel Handling System 1.2-5 1.2.9 Radioactive Waste Management System 1.2-5 1.2.10 Features of Special Interest 1.2-5 1.2.11 Structures 1.2-6 1.3 COMPARISON TABLES 1.3-1 1.3.1 Comparisons with Similar Facility Designs 1.3-1 1.3.2 Comparison of Final and Preliminary Information 1.3-2 1.3.3 References 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Licensee 1.4-1 1.4.2 Architect-Engineer 1.4-1 1.4.3 Reactor Designer 1.4-1 1.4.4 Constructor 1.4-2 1.4.5 Consultants and Service Organization 1.4-2 1.4.5.1 Security Systems - ETA 1.4-2 1.4.5.2 Dames & Moore 1.4-3 1.4.5.3 HARZA Engineering 1.4-3 1.4.5.4 Murray and Trettel, Inc.
1.4-3 1.4.5.5 Shirmer Engineering Corporation 1.4-3 1.4.5.6 Hyla S. Napadensky 1.4-4 1.4.5.7 NALCO Chemical Company 1.4-4 1.4.5.8 Westinghouse Environmental Systems Department (WESD) 1.4-4 1.4.5.9 Illinois Natural History Survey (INHS) 1.4-5 1.4.5.10 NUS Corporation 1.4-5 1.4.5.11 Eberline Instrument Corporation (EIC) 1.4-5


1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Site and Environment 1.2-1 1.2.2 Nuclear Steam Supply System 1.2-1 1.2.3 Engineered Safety Features 1.2-2 1.2.4 Emergency Core Cooling System 1.2-3 1.2.5 Control and Instrumentation 1.2-3 1.2.6 Electrical System 1.2-4 1.2.7 Turbine and Auxiliaries 1.2-4 1.2.8 Fuel Handling System 1.2-5 1.2.9 Radioactive Waste Management System 1.2-5 1.2.10 Features of Special Interest 1.2-5 1.2.11 Structures 1.2-6
B/B-UFSAR 1.0-ii REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS (Cont'd)
PAGE 1.4.5.12 Meteorology Research, Inc. (MRI) 1.4-5 1.4.5.13 Illinois State Museum (ISM) 1.4-6 1.4.5.14 Equitable Environmental Health, Inc. (EEH) 1.4-6 1.4.5.15 Espey, Huston & Associates, Inc. (EH & A) 1.4-6 1.4.5.16 University of Wisconsin-Milwaukee (UWM) 1.4-7 1.4.5.17 Aero-Metric Engineering, Inc. (AME) 1.4-7 1.4.5.18 Iowa Institute of Hydraulic Research 1.4-7 1.4.5.19 Babcock and Wilcox International (B&W) 1.4-8 1.4.5.20 Framatome Technologies, Incorporated (FTI) 1.4-8 1.4.5.21 Stone & Webster Engineers and Constructors, Inc, (S&W) 1.4-8 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Programs Required For Plant Operation 1.5-1 1.5.1.1 Core Stability Evaluation 1.5-1 1.5.2 Other Programs Not Required For Plant Operation 1.5-1 1.5.2.1 Fuel Development Program For Operation at High Power Densities 1.5-2 1.5.2.2 Blowdown Forces Program 1.5-2 1.5.2.3 Blowdown Heat Transfer Testing 1.5-2 1.5.3 References 1.5-4 1.6 MATERIAL INCORPORATED BY REFERENCES 1.6-1 1.7 DRAWINGS 1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Drawings for Independent Structural Review 1.7-1


1.3 COMPARISON TABLES 1.3-1 1.3.1 Comparisons with Similar Facility Designs 1.3-1
B/B-UFSAR 1.0-iii REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF TABLES NUMBER TITLE PAGE 1.3-1 Plants Using Three-Buttress Containment Design 1.3-3 1.4-1 Exelon Generation Company's Nuclear Power Plants in Service or Under Construction 1.4-9 1.4-2 Nuclear Power Plants Completed or Currently Under Design by Sargent & Lundy 1.4-10 1.4-3 Westinghouse Pressurized Water Reactor Nuclear Power Plants 1.4-11 1.5-1 Blowdown Heat Transfer Phase I Test Parameters 1.5-5 1.5-2 Blowdown Heat Transfer Phase II Test Parameters 1.5-6 1.6-1 Topical Reports Incorporated by Reference 1.6-2 1.7-1 Deleted 1.7-2  
 
1.3.2 Comparison of Final and Preliminary Information 1.3-2 1.3.3 References 1.3-2
 
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Licensee 1.4-1 1.4.2 Architect-Engineer 1.4-1 1.4.3 Reactor Designer 1.4-1 1.4.4 Constructor 1.4-2 1.4.5 Consultants and Service Organization 1.4-2 1.4.5.1 Security Systems - ETA 1.4-2 1.4.5.2 Dames & Moore 1.4-3 1.4.5.3 HARZA Engineering 1.4-3 1.4.5.4 Murray and Trettel, Inc. 1.4-3 1.4.5.5 Shirmer Engineering Corporation 1.4-3 1.4.5.6 Hyla S. Napadensky 1.4-4 1.4.5.7 NALCO Chemical Company 1.4-4 1.4.5.8 Westinghouse Environmental Systems Department (WESD) 1.4-4 1.4.5.9 Illinois Natural History Survey (INHS) 1.4-5 1.4.5.10 NUS Corporation 1.4-5 1.4.5.11 Eberline Instrument Corporation (EIC) 1.4-5
 
B/B-UFSAR 1.0-ii  REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS  (Cont'd)
 
PAGE 1.4.5.12 Meteorology Research, Inc. (MRI) 1.4-5 1.4.5.13 Illinois State Museum (ISM) 1.4-6 1.4.5.14 Equitable Environmental Health, Inc. (EEH) 1.4-6 1.4.5.15 Espey, Huston & Associates, Inc. (EH & A) 1.4-6 1.4.5.16 University of Wisconsin-Milwaukee (UWM) 1.4-7 1.4.5.17 Aero-Metric Engineering, Inc. (AME) 1.4-7 1.4.5.18 Iowa Institute of Hydraulic Research 1.4-7 1.4.5.19 Babcock and Wilcox International (B&W) 1.4-8 1.4.5.20 Framatome Technologies, Incorporated (FTI) 1.4-8 1.4.5.21 Stone & Webster Engineers and Constructors, Inc, (S&W) 1.4-8 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Programs Required For Plant Operation 1.5-1 1.5.1.1 Core Stability Evaluation 1.5-1 1.5.2 Other Programs Not Required For Plant Operation 1.5-1 1.5.2.1 Fuel Development Program For Operation at High Power Densities  1.5-2 1.5.2.2 Blowdown Forces Program 1.5-2 1.5.2.3 Blowdown Heat Transfer Testing 1.5-2 1.5.3 References 1.5-4
 
1.6 MATERIAL INCORPORATED BY REFERENCES 1.6-1 
 
===1.7 DRAWINGS===
1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Drawings for Independent Structural Review 1.7-1
 
B/B-UFSAR 1.0-iii REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF TABLES
 
NUMBER TITLE PAGE 1.3-1 Plants Using Three-Buttress Containment Design 1.3-3 1.4-1 Exelon Generation Company's Nuclear Power Plants in Service or Under Construction 1.4-9 1.4-2 Nuclear Power Plants Completed or Currently Under Design by Sargent & Lundy 1.4-10 1.4-3 Westinghouse Pressurized Water Reactor Nuclear Power Plants 1.4-11 1.5-1 Blowdown Heat Transfer Phase I Test Parameters 1.5-5 1.5-2 Blowdown Heat Transfer Phase II Test Parameters 1.5-6 1.6-1 Topical Reports Incorporated by Reference 1.6-2 1.7-1 Deleted 1.7-2  
 
B/B-UFSAR 1.0-iv  REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 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.


B/B-UFSAR 1.0-iv REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 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.
DRAWINGS*
DRAWINGS*
SUBJECT   M-1 General Site Plan Units 1 & 2 M-2 Property Development Units 1 & 2 M-5 General Arrangement Roof Plan Units 1 & 2 M-6 General Arrangement Main Floor At El. 451'-0" Units 1  
SUBJECT M-1 General Site Plan Units 1 & 2 M-2 Property Development Units 1 & 2 M-5 General Arrangement Roof Plan Units 1 & 2 M-6 General Arrangement Main Floor At El. 451-0 Units 1  
& 2 M-7 General Arrangement Mezzanine Floor At El. 426'-0" Units 1 & 2 M-8 General Arrangement Grade Floor At El. 401'-0" Units 1 & 2 M-9 General Arrangement Floor Plan At El. 383'-0" Units 1  
& 2 M-7 General Arrangement Mezzanine Floor At El. 426-0 Units 1 & 2 M-8 General Arrangement Grade Floor At El. 401-0 Units 1 & 2 M-9 General Arrangement Floor Plan At El. 383-0 Units 1  
& 2 M-10 General Arrangement Basement Floor At El. 364'-0" Units 1 & 2 M-11 General Arrangement Floor Plan At El. 346'-0" Units 1  
& 2 M-10 General Arrangement Basement Floor At El. 364-0 Units 1 & 2 M-11 General Arrangement Floor Plan At El. 346-0 Units 1  
& 2 M-12 General Arrangement Radwaste/Service Building Units 1  
& 2 M-12 General Arrangement Radwaste/Service Building Units 1  
& 2 M-13 General Arrangement Fuel Handling Building Units 1 &
& 2 M-13 General Arrangement Fuel Handling Building Units 1 &
2 M-14 General Arrangement Section "A-A" Units 1 & 2 M-15 General Arrangement Section "B-B" Units 1 & 2 M-16 General Arrangement Section "C-C" and "D-D" Units 1 &
2 M-14 General Arrangement Section A-A Units 1 & 2 M-15 General Arrangement Section B-B Units 1 & 2 M-16 General Arrangement Section C-C and D-D Units 1 &
2 M-17 General Arrangement Section "E-E" Units 1 & 2 M-18 General Arrangement Section "F-F" Units 1 & 2 M-19 General Arrangement Lake Screen House Units 1 & 2 (Braidwood)
2 M-17 General Arrangement Section E-E Units 1 & 2 M-18 General Arrangement Section F-F Units 1 & 2 M-19 General Arrangement Lake Screen House Units 1 & 2 (Braidwood)
M-20 General Arrangement River Screen House Units 1 & 2 M-22 General Arrangement Miscellaneous Plans Units 1 & 2 M-34 P&ID Index and Symbols Units 1 & 2  
M-20 General Arrangement River Screen House Units 1 & 2 M-22 General Arrangement Miscellaneous Plans Units 1 & 2 M-34 P&ID Index and Symbols Units 1 & 2  


B/B-UFSAR 1.3-1 REVISION 8 - DECEMBER 2000 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs
B/B-UFSAR 1.1-1 REVISION 15 - DECEMBER 2014 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT


The design is conceptual ly similar to Exelon Generation Company's Zion Station. D ifferences in the design of the two plants have been allowed only (1) when dic tated by the site characteristics, (2) when the change would result in significant safety improvement, simplif ication of constru ction or operation procedures, or cost savi ngs; or (3) as required to comply with appropriate codes and st andards, NRC criteria, regulatory guides, and regulations.  
==1.1 INTRODUCTION==
 
The Nuclear Regulatory Commission approved the transfer of the facility licenses from Commonwealth Edison (ComEd) Company to Exelon Generation Company, LLC (EGC) on August 3, 2000 (Reference 1). References in the Updated Final Safety Analysis Report (UFSAR) to ComEd, CECo, and Commonwealth Edison have been retained, as appropriate, instead of being changed to EGC to properly preserve the historical context.
The nuclear steam supply system is similar to that of the Zion Station but has a slight ly higher power rati ng. The reactor containments are of the same materials and s ize as those at the Zion Station, but each has only three buttresses, rather than six as used at Zion.
This UFSAR is submitted by Exelon Generation Company for nuclear power plants at Byron, Illinois and at Braidwood, Illinois (Drawings M-1 and M-2) in accordance with the requirements of 10 CFR 50.71(e). Each power plant consists of two units having nearly identical nuclear steam supply systems (NSSS) and turbine generators. The main exception is that the original Unit 1 steam generators were replaced by steam generators of a different design. The power plants at the two sites are as nearly identical as site characteristics permit. The bulk of this UFSAR applies to the standardized, non-site-related aspects of the power plants. Sections which describe features specific to the sites are repeated for each site and the applicable station name appears at the top of these pages. Every effort has been made in the preparation of this document to conform to the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", Revision 2, September 1975. The guidance provided in Nuclear Energy Institute (NEI) 98-03, Guidelines for Updating Final Safety Analysis Reports, Revision 1, June 1999, as endorsed by NRC Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10CFR50.71(e), Revision 0, September 1999, is used to comply with the provisions of 10CFR50.71(e).
The number of post-t ensioning tendons is reduced, and the number of wires per tendon incr eased, from that used at Zion. The r educed number of buttresses allows for greater separation of penetration ar eas for redundant safety-related systems.
Each nuclear power plant consists of two nearly identical generating units, and two pressurized water reactor (PWR) (NSSS) and turbine-generator furnished by Westinghouse Electric Corporation (Westinghouse) similar in design concept to several projects recently licensed or currently under review by the NRC (see Section 1.3). Unit 1 contains steam generators supplied by B&W and Unit 2 contains steam generators supplied by Westinghouse.
Several plants on which this buttress design has been used are listed in Table 1.3-1.
Westinghouse Electric Corporation, Sargent &
The polar cranes in the reactor containment are mounted on the containment wall, rather than on the missile barrier as at Zion.
Lundy, and the Commonwealth Edison Company jointly participated in the original design and construction of each unit. The plant is operated by Exelon Generation Company. Sargent & Lundy (S&L) is the architect-engineer for both stations.
This allows use of a g reater area for co mponent laydown in the containment.  
Each nuclear steam supply system (NSSS) has been evaluated at a power output of 3672 MWt for the Measurement Uncertainty Recapture (MUR) Power Uprate. The warranted gross and approximate net electrical outputs for the MUR are 1268 MWe and 1241 MWe for Unit 1 and Unit 2, respectively. Safety analyses are evaluated at an NSSS power level of 3672 MWt and a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level of 3648 MWt.


Two 100%-capacity cont ainment spray systems are used, rather than the three systems used at Zion.
B/B-UFSAR 1.1-1a REVISION 15 - DECEMBER 2014 Specifically, the containment and engineered safety features (ESF) are designed and evaluated for operation at a core thermal power level of 3658 MWt. Accidents (such as loss-of-coolant, steamline break, and other postulated accidents having offsite dose consequences) are also analyzed at a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level 3648 MWt.
Four containment fan coolers are used, rather than the five u sed at Zion.
The emergency diesel-generator systems for each unit are entir ely independent and use two 5500-kW diesel generators per unit.
The arrangement of equipment in the co mmon auxiliary bui lding allows greater physical separation of r edundant systems and their piping and


cables than was possible at Zion.
B/B-UFSAR 1.1-2 REVISION 9 - DECEMBER 2002 The reactor containments are of post-tensioned concrete construction with a carbon steel liner. Sufficient free volume is provided to contain the energy released in a major accident without need for "pressure suppression" devices. Sargent & Lundy is responsible for containment design.
The Byron Station uses natural draft cooling towers for heat rejection. Zion utilizes once-t hrough cooling.
Byron Station is located in north central Illinois, near the town of Byron and near the Rock River (Drawing M-1). Cooling for the plant is provided by two natural draft cooling towers for non-essential service cooling, and by mechanical draft cooling towers for essential cooling. The fuel loading dates for the two units were November 1984 and November 1986 for Units 1 and 2, respectively. The corresponding dates for commercia1 operation were September 1985 and August 1987.
Mechanical draft cooling towers are provided for essential service cooling at Byron.  
The Braidwood Station is located in northeastern Illinois, near the town of Braidwood and near the Kankakee River (Drawing M-1).
Cooling for the plant is provided by a large man-made cooling pond of approximately 2500 acres constructed over a previously strip-mined area. Essential service cooling is provided by a 99-acre auxiliary cooling pond which is integral with the main pond. The fuel loading dates for the two units were October 1986 and December 1987 for Units 1 and 2, respectively. The corresponding dates for commercial operation were July 1988 and October 1988.
The standard symbols used on piping and instrument diagrams and other figures in this UFSAR are shown in Drawing M-34.


The Braidwood Statio n uses a large man-made cooling pond for heat rejection. An auxiliary cooling pond, integral with the main pond, is provided for es sential service cooling.
==1.2 REFERENCES==
1.
NRC letter, "Braidwood, Byron, Dresden, LaSalle, Quad Cities, and Zion - Orders Approving Transfer of Licenses From Commonwealth Edison Company To Exelon Generation Company, LLC, and Approving Conforming Amendments," dated August 3, 2000


Table 1.3-2 of the FSAR provided the des ign comparison of the Byron/Braidwood nuclear steam supply system wi th Comanche Peak, Indian Point 2, South Texas, Sun Deser t, W. B. McGuire Nuclear Station, Trojan Nuclear Power Plant, SNUPPS, and the Watts Bar Application. This inf ormation was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.
B/B-UFSAR 1.2-1 REVISION 15 - DECEMBER 2014 1.2 GENERAL PLANT DESCRIPTION 1.2.1 Site and Environment The characteristics of the sites and their environs have been investigated to establish bases for determining criteria for storm, flood, and earthquake protection and to evaluate the validity of calculational techniques for the control of routine and accidental releases of radioactive liquids and gases to the environment. Field programs to investigate geology and seismology are completed. Preoperational meteorological programs to provide onsite observations of wind speed and direction have continued since the spring of 1973 at Byron and since the fall of 1973 for Braidwood. Radiological studies of the site environs were initiated at least 18 months prior to commercial operation, with the objective of establishing background radiation levels.
B/B-UFSAR 1.3-2 REVISION 8 - DECEMBER 2000 1.3.2 Comparison of Final and Preliminary Information The Byron/Braidwood Power Plant design was s ubject to continuing review throughout the construction of the stations. The experience gained at Zion Stat ion and other PWRs was used to enhance equipment reli ability and perfor mance. Current design technology was used to upgrade earlier p lant design methods.
The geography, demography, meteorology, hydrology, geology, and seismology of the two plant sites are discussed in detail in Chapter 2.0.
No significant design ch anges have been made to the Byron Station or the Braidwood Stati on which have not been previously reported by amendment to the PS AR, except for the inclusion of 17 x 17
1.2.2 Nuclear Steam Supply System The nuclear steam supply system (NSSS) consists of a Westinghouse pressurized water reactor and supporting auxiliary systems.
Performance at the calculated steam flow of the NSSS at MUR conditions based on zero percent makeup is as follows:
a.
thermal output of NSSS (MWt) - 3659; b.
thermal output of reactor core (MWt) -3645; c.
steam flow from NSSS (lb/hr) - 16,347,514 for Unit 1/16,280,677 for Unit 2; d.
steam pressure at a steam generator outlet (psia) -
1020.8 for Unit 1 and 902 for Unit 2; e.
maximum moisture content (%) - 0.25%; and f.
feedwater temperature at steam generator inlet (F) -
446.5 for Unit 1 and 447.5 for Unit 2.
The NSSS consists of a reactor and closed reactor coolant loops connected in parallel to the reactor vessel, each loop containing a reactor coolant pump and a steam generator. The NSSS also contains an electrically heated pressurizer and certain auxiliary systems.


optimized fuel. Table 1.3-3 of the FSAR lis ted those significant changes reported since t he issuance of the B yron and Braidwood Stations Construction Pe rmits. This informati on was current at the time the Byron U nit 1 operating lice nse was granted and has not been included in the UFSAR.  
B/B-UFSAR 1.2-1a REVISION 7 - DECEMBER 1998 High pressure reactor coolant circulates through the reactor core to remove the heat generated by the nuclear reaction. The heated reactor coolant flows from the reactor vessel to the steam generators (via reactor coolant loop piping). The coolant gives up its heat to the feedwater in the steam generator to generate steam for the turbine generator. The cycle is completed when the reactor coolant is pumped back to the reactor vessel. The entire reactor coolant system is composed of leaktight components to contain the reactor coolant to the system.


Other changes included t he removal of the part length control rods (they are not n eeded to control X enon induced axial instabilities), the enla rgement of spent fue l capacity, the use of more corrosion-resist ant materials in the s team generators and moisture steam separator s, improved equipment packaging to do a reactor refueling in a shorter time peri od, an upgraded design for the reactor coolant pump seals, and replacement steam generators for Unit 1. These concepts are described in later chapters.  
B/B-UFSAR 1.2-2 REVISION 11 - DECEMBER 2006 The core is a multiregion type. All fuel assemblies are mechanically identical, although the fuel enrichment is not the same in all assemblies. In a typical initial core loading, three fuel enrichments are used in mechanically identical assemblies.
Fuel assemblies with the highest enrichments are placed in the core periphery, or outer region, and the two groups of lower enrichment fuel assemblies are arranged in a selected pattern in the central region. In subsequent refuelings, one third of the fuel is discharged and fresh fuel is loaded into the outer region of the core. The remaining fuel is arranged in the central two-thirds of the core in such a manner as to achieve optimum power distribution.
Rod cluster control assemblies are used for reactor control and consist of clusters of cylindrical absorber rods. The absorber rods move within guide tubes in certain fuel assemblies. Above the core, each cluster of absorber rods is attached to a spider connector and drive shaft, which is raised and lowered by a drive mechanism mounted on the reactor vessel head. The insertion of the rod cluster control assembly for a reactor trip is by gravity.
The reactor coolant pumps are Westinghouse vertical, single-stage, centrifugal pumps of the shaft-seal type.
The steam generators are B&W vertical U-tube units for Unit 1 and Westinghouse vertical U-tube units for Unit 2. All steam generators contain Inconel tubes. Integral moisture separation equipment reduces the moisture content of the steam.
The reactor coolant piping and all of the pressure-containing surfaces in contact with reactor water are stainless steel. The steam generator tubes and fuel cladding are Inconel and Zircaloy/ZIRLO, respectively. Reactor core internals, including control rod drive shafts, are primarily stainless steel.
An electrically heated pressurizer connected to one reactor coolant loop maintains reactor coolant system pressure during normal operation, limits pressure variations during plant load transients, and keeps system pressure within design limits during abnormal conditions.
Auxiliary system components are provided to charge makeup water into the reactor coolant system, purify reactor coolant, provide chemicals for corrosion inhibition and reactivity control, cool system components, remove decay heat, and provide for emergency safety injection.
1.2.3 Engineered Safety Features The engineered safety features provided for this plant have sufficient redundancy of components and power sources such that


====1.3.3 References====
B/B-UFSAR 1.2-3 REVISION 12 - DECEMBER 2008 under the conditions of a loss-of-coolant accident they can maintain the containment integrity and limit personnel exposure to less than 10 CFR 50.67 limits. The engineered safety features incorporated in the design of this plant and the functions they serve are summarized in the following.
: 1. Exelon Generatio n Company, "Byron/Br aidwood Stations Fire Protection Report in Response to Appendix A of BTP APCSB 9.5-1,"
a.
(current amendment).  
The emergency core cooling system injects borated water into the reactor coolant system if coolant is lost. This system limits damage to the core and limits the fission product contamination released into the containment following a postulated loss-of-coolant accident (LOCA).
b.
A steel lined, concrete containment vessel consists of a post-tensioned concrete cylindrical wall and shallow dome, and a conventionally reinforced concrete base. The containment forms a virtually leaktight barrier to prevent the escape of radioactivity.
c.
Reactor containment fan coolers reduce containment temperature and pressure following a postulated loss-of-coolant accident.
d.
A containment spray system is used to reduce containment pressure and to remove iodine and particulate fission products from the containment atmosphere in the event of a loss-of-coolant accident.
e.
The auxiliary feedwater system provides for heat removal from the reactor coolant system by providing makeup water to the steam generator under a variety of postulated conditions.
f.
A combustible gas control system is provided to ensure that the containment atmosphere is mixed following a loss-of-coolant accident. A mixed containment atmosphere prevents local accumulation of combustible or detonable gases that could threaten containment integrity or equipment operating in a local compartment.
1.2.4 Emergency Core Cooling System The emergency core cooling system (ECCS), with passive and active subsystems, is designed to inject borated water into the reactor coolant system (RCS) following a LOCA. This will provide cooling to limit core damage, metal-water reactions, and fission-product release. The ECCS provides long-term postaccident cooling of the core by drawing borated water from the containment sump.
1.2.5 Control and Instrumentation The reactor is controlled by a variety of reactivity coefficients (temperature, pressure, doppler) by control rod cluster motion which is required for load follow transients and for startup and shutdown, and by a soluble neutron absorber, i.e., boron in the


B/B-UFSAR 1.3-3 TABLE 1.3-1 PLANTS USING T HREE-BUTTRESS CO NTAINMENT DESIGN
B/B-UFSAR 1.2-4 REVISION 15 - DECEMBER 2014 form of boric acid which is adjusted in concentration to compensate for such effects as fuel consumption and accumulation of fission products.
1.2.6 Electrical System Each unit's main generator is an 1800-rpm, 3-phase, 60-cycle, hydrogen-innercooled unit with water-cooled stator windings and is rated at 1361 MVA at 75 psig gas pressure. Field excitation is provided by a direct shaft-driven brushless excitation system. Two one-half size main step-up transformers deliver power to the 345-kV switchyard.
The station's auxiliary power system consists of system and unit auxiliary transformers; 6900-V, 4160-V, and 480-V switchgear; 480-V motor control centers; 120-Vac instrument buses; and 250-Vdc and 125-Vdc buses.
Two diesel generators are provided for each unit and are available as onsite sources of power (in the event of complete loss of normal a-c power) for operating essential safeguard features. Each diesel generator is capable of supplying required electrical loads for a simultaneous LOCA and loss of offsite power to any one unit.
1.2.7 Turbine and Auxiliaries The turbine for each unit is a four-casing, tandem-compound, six-flow exhaust, 1800-rpm unit with 40-inch last-row blades. There are two combination moisture-separator/steam-reheater assemblies per unit. The turbine-generator for Units 1 have a MUR rating of 1268 MWe gross at 16,347,514 lb/hr steam flow with inlet steam conditions of 1001 psia, 0.36% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. The turbine-generators for Units 2 have a MUR rating of 1241 MWe gross at 16,280,677 lb/hr steam flow with inlet steam conditions of 882 psi, 0.34% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. There are seven stages of feedwater heating for all units.
The turbine is equipped with a redundant fault tolerant Westinghouse Ovation based distributed control system. All control algorithms and processes within the turbine control system are redundant and configured to allow unrestricted turbine operation. This system utilizes a fire-resistant hydraulic fluid to control throttle and governor valve positioning.


PLANT/UTILITY DATE OF OPERATION
B/B-UFSAR 1.2-4a REVISION 11 - DECEMBER 2006 The condenser is of the single-pass deaerating type. There are three parallel strings of feedwater heaters that utilize extraction steam from the low pressure turbines, two parallel strings of feedwater heaters that utilize extraction and exhaust steam from the high pressure turbine, four one-third-sized feedwater condensate and condensate booster pumps, and three one-half-sized feedwater and heater drain pumps. Heater drains from the three highest-pressure feedwater heaters are pumped into the feedwater system; drains from the four lowest-pressure heaters are cascaded to the condenser.


Arkansas Nuclear One
B/B-UFSAR 1.2-5 REVISION 14 - DECEMBER 2012 1.2.8 Fuel Handling System The reactor is refueled with equipment which handles the spent fuel under water from the entire time from leaving the reactor vessel until it is secured in a cask for shipment. Underwater transfer of spent fuel provides a transparent radiation shield and a reliable coolant for decay heat removal.
Fuel handling is performed in the refueling cavity which is flooded for refueling, and the fuel storage pool which is in the fuel building. The two areas are connected by a fuel transfer system which carries the fuel through an opening in the reactor containment.
Spent fuel is removed from the reactor vessel by a refueling machine, placed on the fuel transfer cart conveyor and transferred to the spent fuel storage pool. The fuel is removed from the transfer cart and placed into storage racks. After a suitable decay period, the fuel may be removed from storage and loaded into a shipping cask for removal from the plant.
Refer to Section 9.1.2.3.11 for a description of spent fuel storage and handling using Dry Cask Storage (DCS) system and the Independent Spent Fuel Storage Installation (ISFSI).
All important pumps, piping, and equipment are replicated and capable of being supplied from one of two independent ESF divisions.
1.2.9 Radioactive Waste Management System The radioactive waste system provides equipment necessary to collect, process, and prepare for the disposal of radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation or to transfer the wastes to a vendor-supplied radwaste system.
After collection, depending on chemical composition, liquid wastes may be demineralized and/or filtered. The treated water is discharged at concentrations within the limits of 10 CFR 20.
Sludges and spent demineralizer resins are processed by a vendor-supplied radwaste system for ultimate disposal in an authorized location.
Gaseous wastes are collected from the waste gas header.
Discharge of the gaseous wastes to the environment is controlled to ensure that the offsite dose is as low as reasonably achievable (ALARA).
1.2.10 Features of Special Interest The fundamental concept for the design and construction of the Byron/Braidwood Stations is one of commonality and duplication to the maximum extent permitted by site characteristics. For those features not dictated specifically by site characteristics, identical designs have been employed for the two stations. The concept has been extended to the point where the limiting (i.e.,


Arkansas Power & Light Co.
B/B-UFSAR 1.2-6 REVISION 9 - DECEMBER 2002 worst case) parameters of the sites are considered in the common design. An example of this is the use of the most restrictive site's seismic building response spectra for the design of systems and components in both plants.
5-21-74 Millstone-2
Common plans, drawings, and specifications have been issued for construction at the two sites. Design and construction management for both sites have been conducted by the same major organizations, using the same quality assurance and project management programs. This approach embraces the concept of standardization in nuclear power plant design and construction.
1.2.11 Structures The major structures include a separate and independent containment for each reactor, a common auxiliary building, a common turbine building, a common solid radwaste storage, and administration and service building. General layouts of the plant and interior components' arrangements are shown on Drawings M-5 through M-18 and M-20 and M-22 (Byron), and Drawings M-5 through M-20 and M-22 (Braidwood).
For purposes of design and analysis, structures are designated by Safety Category according to their relation to plant safety. The Safety Category definitions are as follows:
a.
Safety Category I - Those structures important to safety that must be designed to remain functional in the event of the safe shutdown earthquake (SSE) and other design-basis events (including tornado, probable maximum flood, operating basis earthquake (OBE), missile impact, or accident internal to the plant) are designated as Safety Category I.
b.
Safety Category II - Those structures which are not designated as Safety Category I are designated as Safety Category II.
The design criteria and analysis methods for these structures are discussed in Chapter 3.0.


Northeast Utilities 8-1-75 Rancho Seco Sacramento Municipal Utility District 8-16-74  Trojan Portland General Electric Co.
B/B-UFSAR 1.3-1 REVISION 8 - DECEMBER 2000 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs The design is conceptually similar to Exelon Generation Company's Zion Station. Differences in the design of the two plants have been allowed only (1) when dictated by the site characteristics, (2) when the change would result in significant safety improvement, simplification of construction or operation procedures, or cost savings; or (3) as required to comply with appropriate codes and standards, NRC criteria, regulatory guides, and regulations.
11-21-75
The nuclear steam supply system is similar to that of the Zion Station but has a slightly higher power rating. The reactor containments are of the same materials and size as those at the Zion Station, but each has only three buttresses, rather than six as used at Zion. The number of post-tensioning tendons is reduced, and the number of wires per tendon increased, from that used at Zion. The reduced number of buttresses allows for greater separation of penetration areas for redundant safety-related systems. Several plants on which this buttress design has been used are listed in Table 1.3-1.
The polar cranes in the reactor containment are mounted on the containment wall, rather than on the missile barrier as at Zion.
This allows use of a greater area for component laydown in the containment.
Two 100%-capacity containment spray systems are used, rather than the three systems used at Zion. Four containment fan coolers are used, rather than the five used at Zion. The emergency diesel-generator systems for each unit are entirely independent and use two 5500-kW diesel generators per unit. The arrangement of equipment in the common auxiliary building allows greater physical separation of redundant systems and their piping and cables than was possible at Zion.
The Byron Station uses natural draft cooling towers for heat rejection. Zion utilizes once-through cooling. Mechanical draft cooling towers are provided for essential service cooling at Byron.
The Braidwood Station uses a large man-made cooling pond for heat rejection. An auxiliary cooling pond, integral with the main pond, is provided for essential service cooling.
Table 1.3-2 of the FSAR provided the design comparison of the Byron/Braidwood nuclear steam supply system with Comanche Peak, Indian Point 2, South Texas, Sun Desert, W. B. McGuire Nuclear Station, Trojan Nuclear Power Plant, SNUPPS, and the Watts Bar Application. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.


J.M. Farley-1
B/B-UFSAR 1.3-2 REVISION 8 - DECEMBER 2000 1.3.2 Comparison of Final and Preliminary Information The Byron/Braidwood Power Plant design was subject to continuing review throughout the construction of the stations. The experience gained at Zion Station and other PWRs was used to enhance equipment reliability and performance. Current design technology was used to upgrade earlier plant design methods.
 
No significant design changes have been made to the Byron Station or the Braidwood Station which have not been previously reported by amendment to the PSAR, except for the inclusion of 17 x 17 optimized fuel. Table 1.3-3 of the FSAR listed those significant changes reported since the issuance of the Byron and Braidwood Stations Construction Permits. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.
Alabama Power Co.
Other changes included the removal of the part length control rods (they are not needed to control Xenon induced axial instabilities), the enlargement of spent fuel capacity, the use of more corrosion-resistant materials in the steam generators and moisture steam separators, improved equipment packaging to do a reactor refueling in a shorter time period, an upgraded design for the reactor coolant pump seals, and replacement steam generators for Unit 1. These concepts are described in later chapters.
6-25-77 B/B-UFSAR 1.4-1 REVISION 8 - DECEMBER 2000 1.4 IDENTIFICATION OF AG ENTS AND CONTRACTORS
1.3.3 References
: 1. Exelon Generation Company, "Byron/Braidwood Stations Fire Protection Report in Response to Appendix A of BTP APCSB 9.5-1,"
(current amendment).


====1.4.1 Licensee====
B/B-UFSAR 1.3-3 TABLE 1.3-1 PLANTS USING THREE-BUTTRESS CONTAINMENT DESIGN PLANT/UTILITY DATE OF OPERATION Arkansas Nuclear One Arkansas Power & Light Co.
5-21-74 Millstone-2 Northeast Utilities 8-1-75 Rancho Seco Sacramento Municipal Utility District 8-16-74 Trojan Portland General Electric Co.
11-21-75 J.M. Farley-1 Alabama Power Co.
6-25-77


Exelon Generation Company is the Licensee for the Byron Station, Units 1 and 2, which is located in Roc kvale Township, Ogle County, approximately 4 miles south of Byron
B/B-UFSAR 1.4-1 REVISION 8 - DECEMBER 2000 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 Licensee Exelon Generation Company is the Licensee for the Byron Station, Units 1 and 2, which is located in Rockvale Township, Ogle County, approximately 4 miles south of Byron, Illinois, and for Units 1 and 2 of the Braidwood Station, which is located in Reed Township, Will County, approximately 6 miles southwest of Wilmington, Illinois. The Licensee is responsible for the design, construction, and operation of the nuclear power plants.
, Illinois, and for Units 1 and 2 of the B raidwood Station, whic h is located in Reed Township, Will County, approximately 6 m iles southwest of Wilmington, Illinois.
Commonwealth Edison supplies electrical service to an area of 13,000 square miles with a population of approximately 8 million persons, located primarily in the northern third of Illinois.
The Licensee is r esponsible for the design, construction, and operation of the n uclear power plants.  
Dresden 1, Commonwealth Edison's first nuclear generating station, went into commercial service during August 1960, and has produced more than 10 billion kWh. Additional nuclear units in service or under construction are listed in Table 1.4-1.
1.4.2 Architect-Enqineer For the work covered by this application, Sargent & Lundy (S&L) has been retained as the design consultants. The Licensee has employed Sargent & Lundy for power plant design work for over 80 years.
Sargent & Lundy is an independent consulting engineering organization founded in Chicago, in 1891. For over three-quarters of a century, the firm has specialized exclusively in the design of generation, transmission, distribution, and utilization of steam and electric power and related facilities.
The firm has provided the complete engineering services for more than 600 turbine-generator units with a total capacity of 53,000,000 kW. Of this total, some 9,800,000 kW is in nuclear generating capacity. Table 1.4-2 lists the nuclear plants completed by or currently under design by Sargent & Lundy.
1.4.3 Reactor Designer Westinghouse has designed, developed, and manufactured nuclear power facilities since the 1950s, beginning with the world's first large central station nuclear power plant (Shippingport),
which started producing power in 1957. Completed or contracted


Commonwealth Edison su pplies electrical serv ice to an area of 13,000 square miles with a population of app roximately 8 million persons, located primarily in the northern thi rd of Illinois.  
B/B-UFSAR 1.4-2 REVISION 5 - DECEMBER 1994 commercial nuclear capacity totals were in excess of 98,000 MWe. Westinghouse pioneered new nuclear design concepts, such as chemical shim control of reactivity and the rod cluster control concept, throughout the last two decades. Westinghouse manufacturing facilities include the largest commercial nuclear fuel fabrication facility in the world and the world's most modern heat transfer equipment production facility, as well as other facilities producing nuclear steam supply system (NSSS) components. Table 1.4-3 lists all Westinghouse pressurized water reactor (PWR) plants to date, including those plants under construction or on order at the time of the Byron/Braidwood application.
The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute have contracted with Westinghouse for research into NSSS-related activities. Westinghouse experience was also utilized by the NRC and Metropolitan Edison immediately following the Three Mile Island Unit 2 accident and the corporation continues to participate with the Westinghouse Owner's Group of utilities in addressing the NRC action plan and other operations improvements.
1.4.4 Constructor Construction coordination of all activities at the site was under the supervision of the Commonwealth Edison's Station Construction Department. The department exercises site managerial functions as discussed in Chapter 17.0 of the UFSAR.
The Station Construction Department was the constructor for Zion Station. This department has coordinated the construction activities for almost all of Commonwealth Edison's existing power plants. It was also the construction coordinator for La Salle County Station.
1.4.5 Consultants and Service Organization 1.4.5.1 Security System - ETA The design of the physical security system and the administrative controls was performed by ETA, Inc.
ETA personnel have had varied and in-depth experience in the design, safety analysis, and environmental review of nuclear power plants and related facilities as well as in the management and organization of security systems. They are very familiar with the details of the current generation of light water reactors and, in particular, those critical areas and components of the plants which might be the most vulnerable to sabotage.
They are also familiar with the current regulations and guidelines of the NRC that define the required performance and objectives of a security system for licensed activities.  


Dresden 1, Commonwea lth Edison's first n uclear generating station, went into comme rcial service during A ugust 1960, and has produced more than 10 billion kWh.
B/B-UFSAR 1.4-3 1.4.5.2 Dames & Moore The independent consulting firm of Dames & Moore was employed to conduct studies relating to the geology, seismology, and groundwater hydrology at both sites. The firm also conducted preconstruction baseline studies, including wildlife surveys as well as soil and vegetation analyses.
Additional nucle ar units in service or under con struction are listed in Table 1.4-1.
Having performed environmental studies for approximately 30 nuclear power plant sites, Dames & Moore is a recognized authority in the field of environmental engineering of nuclear power plants.
 
1.4.5.3 HARZA Engineering HARZA was employed in the design of the water treatment facilities at both stations.
1.4.2  Architect-Enqineer
HARZA has been involved with a variety of technical studies for at least ten nuclear power stations. Among these studies have been facility design, review of design and structure, hydrology, and groundwater. In addition, HARZA Engineering has designed some of the largest hydroelectric projects in the world, including major concrete structures and earthfilled dams.
 
For the work covered by this application, Sa rgent & Lundy (S&L) has been retained as the design consultants. The Licensee has employed Sargent & Lundy for pow er plant design work for over 80 years. Sargent & Lundy is an independent cons ulting engineering organization founded in Chicago, in 1891.
For over three-quarters of a century, the firm has spec ialized exclusively in the design of generation, tr ansmission, dist ribution, and utilization of steam a nd electric power and related facilities.
The firm has provided the comple te engineering s ervices for more than 600 turbine-generator u nits with a total capacity of 53,000,000 kW. Of t his total, some 9,80 0,000 kW is in nuclear generating capacity.
Table 1.4-2 lists the nuclear plants completed by or curr ently under design by Sargent & Lundy.
1.4.3  Reactor Designer
 
Westinghouse has designed, dev eloped, and manu factured nuclear power facilities since t he 1950s, beginning with the world's first large central stat ion nuclear power pl ant (Shippingport),
which started producing power in 1957. Comple ted or contracted
 
B/B-UFSAR 1.4-2 REVISION 5 - DECEMBER 1994 commercial nuclear cap acity totals were in excess of 98,000 MWe. Westinghouse p ioneered new nuclear design concepts, such as chemical shim control of reac tivity and the rod cluster control concept, through out the last two decad es. Westinghouse manufacturing facilities include the largest com mercial nuclear fuel fabrication facility in the world and t he world's most modern heat transfer e quipment production facility, as well as other facilities producing nucle ar steam supply system (NSSS) components. Table 1
.4-3 lists all Wes tinghouse pressurized water reactor (PWR) pl ants to date, incl uding those plants under construction or on ord er at the time of the Byron/Braidwood application.
The U.S. Nuclear Regulatory Commissi on (NRC) and the Electric Power Research I nstitute have contracted with Westinghouse for research into NSSS-related activ ities. Westinghouse experience was also utilized by the NRC and Metropolitan Edison immediately following the Three Mile Island Unit 2 accident and the corporation continues to participate wit h the Westinghouse Owner's Group of utilities in ad dressing the NRC action plan and other operations improvements.
 
====1.4.4 Constructor====
Construction coordination of all activities at the site was under the supervision of the C ommonwealth Edison's Station Construction Department. Th e department exercises site managerial functions as discussed in Chapter 1 7.0 of the UFSAR.
The Station Construction Departm ent was the constructor for Zion Station. This departm ent has coordinated the construction activities for almost all of Commonwealth Ed ison's existing power plants. It was also the construction coordi nator for La Salle County Station.
 
1.4.5  Consultants and Service Organization 1.4.5.1 Security System - ETA
 
The design of the physical secur ity system and the administrative controls was perform ed by ETA, Inc.
ETA personnel have had varied and in-depth experience in the design, safety analysis, and environmental r eview of nuclear power plants and related facilit ies as well as in the management and organization of security sys tems. They are very familiar with the details of the current generation of light water reactors and, in particular, those critical areas and components of the plants wh ich might be the most vulnerable to sabotage.
They are also famili ar with the current regulations and guidelines of the NRC that define the required performance and objectives of a security system for licensed activities.
 
B/B-UFSAR 1.4-3 1.4.5.2 Dames & Moore The independent consulti ng firm of Dames &
Moore was employed to conduct studies relating to the geo logy, seismology, and groundwater hydrology at both sites. The firm also conducted preconstruction baseline studies, including wi ldlife surveys as well as soil and vegetation analyses.
Having performed envir onmental studies for approximately 30 nuclear power plant si tes, Dames & Moore is a recognized authority in the field of en vironmental engine ering of nuclear power plants.
1.4.5.3 HARZA Engineering
 
HARZA was employed in the design of the water treatment facilities at both stations.
HARZA has been involved with a variety of technical studies for at least ten nuclear power sta tions. Among these studies have been facility design, review of design and structure, hydrology, and groundwater. In addition, HARZA Eng ineering has designed some of the largest hy droelectric projects in the world, including major concrete struc tures and earthfilled dams.
1.4.5.4 Murray and Trettel, Inc.
1.4.5.4 Murray and Trettel, Inc.
Murray and Trettel (M&T) is an environmental consulting firm which, since 1960, has provided significant meteorological input to both preoperational and operational phases of meteorological programs for nuclear power stations. M&T has also provided meteorological input to a wide variety of air pollution and environmental problems as well as allied control technique programs.
Murray & Trettel provided meteorological data for both stations by implementation of an onsite measurement program incorporating a tower for elevation measurements.
1.4.5.5 Shirmer Engineering Corporation Shirmer Engineering is a firm of consulting fire protection engineers. The firm has done work on 17 Department of Energy nuclear fuel production and laboratory facilities, as well as for numerous nuclear power stations for Sargent & Lundy.
Shirmer Engineering has also performed services for many fossil units.
Shirmer Engineering provided evaluation of the fire protection systems at both stations and assisted in the preparation of the Byron/Braidwood Fire Protection Report.


Murray and Trettel (
B/B-UFSAR 1.4-4 1.4.5.6 Hyla S. Napadensky Ms. Napadensky was retained to help evaluate the probability of an accidental explosion occurring on a train carrying explosives in the vicinity of the Braidwood Station.
M&T) is an environmental consulting firm which, since 1960, has provided signif icant meteorological input to both preoperational and operational phases of meteorological progr ams for nuclear power stat ions. M&T has also provided meteorological input to a wide variety of air pollution and environmental prob lems as well as allied control technique programs.
Ms. Napadensky is the Manager of Fire Safety Research at the IIT Research Institute of the Illinois Institute of Technology.
 
Ms. Napadensky has directed analytical and experimental research in the areas of explosion effects, hazards and risk analysis, safety of chemical systems, explosives and propellant sensitivity, and initiation mechanisms during her 17 years with IIT Research Institute.
Murray & Trettel provided meteor ological data for both stations by implementation of an onsite measurement p rogram incorporating a tower for elev ation measurements.
1.4.5.7 NALCO Chemical Company The NALCO Chemical Company (formerly Industrial Bio-Test, Inc.)
1.4.5.5 Shirmer Engi neering Corporation
consisted of two divisions, Industrial Bio-Test Laboratories, and NALCO Environmental Sciences, which conduct studies relating to toxicology and ecological sciences, respectively. The Environmental Science Division includes seven subdivisions: (1) aquatic biology, (2) fisheries and field operations, (3) water and wastewater chemistry, (4) radiochemistry, (5) air sciences and data processing, (6) land and plant sciences, and (7) environmental physiology.
 
As a technical consultant on the Braidwood project, the NALCO Chemical Company provided a clam bed mapping survey in the area of the station's intake and discharge structures located on the Kankakee River.
Shirmer Engineering is a firm of consulting fire protection engineers. The firm has done work on 17 Department of Energy nuclear fuel production and laboratory facilit ies, as well as for numerous nuclear power stations for Sargent & Lundy.
1.4.5.8 Westinghouse Environmental Systems Department (WESD)
Shirmer Engineering has also performed servi ces for many fossil units. Shirmer Engineering prov ided evaluation of the fire protection systems at both stations and assisted in the preparation of the Byron/Braidwood Fire Protection Report.
WESD, established as a department of the Westinghouse Power Systems Company in 1969, consisted of environmental scientists and engineers experienced in the areas of aquatic and terrestrial biology and ecology; geology; limnology; environmental chemistry and physics; physical oceanography, meteorology and climatology, radiology, public health aspects of pollutant emissions, and systems engineering and integration.
 
WESD conducts broad environmental surveys, environmental program planning and data interpretation, and provides recommended action programs for meeting federal, state, and local environmental quality regulations. As a technical consultant on the Braidwood project, WESD staff biologists conducted a 2-year baseline study of the Braidwood Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Kankakee River in the vicinity of the site were determined, and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation in the biotic communities of the site were predicted.  
B/B-UFSAR 1.4-4 1.4.5.6 Hyla S. Napadensky Ms. Napadensky was ret ained to help evaluate the probability of an accidental explosion occurring on a train carrying explosives in the vicinity of t he Braidwood Station.
Ms. Napadensky is the Manager of Fire Safety Research at the IIT Research Institute of the Illinois Insti tute of Technology.
Ms. Napadensky has dir ected analytical and e xperimental research in the areas of explosion effects, haz ards and r isk analysis, safety of chemical systems, explosives and propellant sensitivity, and initiat ion mechanisms during her 17 years with IIT Research Institute.  
 
1.4.5.7 NALCO Ch emical Company
 
The NALCO Chemical Com pany (formerly Industr ial Bio-Test, Inc.)
consisted of two divisions, Indu strial Bio-Test Labo ratories, and NALCO Environmental Scie nces, which conduct st udies relating to toxicology and ecolo gical sciences, re spectively. The Environmental Science Di vision includes seven subdivisions: (1) aquatic biology, (2) fis heries and field opera tions, (3) water and wastewater chemistry, (4) radiochemistry, (5) air sciences and data processing, (6) land and plant sciences, and (7) environmental physiology.
As a technical consult ant on the Braidwo od project, the NALCO Chemical Company provided a clam bed mapping sur vey in the area of the station's intake and discharge struct ures located on the Kankakee River.
1.4.5.8 Westinghouse E nvironmental Syste ms Department (WESD)
 
WESD, established as a department of t he Westinghouse Power Systems Company in 1969, consisted of environmental scientists and engineers experienced in the areas of aqua tic and terrestrial biology and ecology; g eology; limnology; env ironmental chemistry and physics; physical oceanography, meteorol ogy and climatology, radiology, public health aspec ts of pollutant emissions, and systems engineering and integration.  


WESD conducts broad en vironmental surveys, environmental program planning and data interpretation, and provides recom mended action programs for meeting federal, state, and local environmental quality regulations. As a technical consult ant on the Braidwood project, WESD staff biol ogists conducted a 2-y ear baseline study of the Braidwood Stati on site. Distribution s of phytoplankton, zooplankton, periphyton, benthos, fi sh, fish eggs and larvae, and water chemistry in the Kankakee Rive r in the vicinity of the site were determined, and quantitative data on terr estrial flora and fauna were collected.
The impacts of plant construction and operation in the biotic communit ies of the site were predicted.
B/B-UFSAR 1.4-5 1.4.5.9 Illinois Natural History Survey (INHS)
B/B-UFSAR 1.4-5 1.4.5.9 Illinois Natural History Survey (INHS)
The Illinois Natural History Survey (INH S), which has its beginnings almost 12 0 years ago, is a di vision of the State Department of Registra tion and Education and provides services to farmers, homeowners, sportsmen, and all other citizens of Illinois as well as to industries.
The Illinois Natural History Survey (INHS), which has its beginnings almost 120 years ago, is a division of the State Department of Registration and Education and provides services to farmers, homeowners, sportsmen, and all other citizens of Illinois as well as to industries. INHS cooperates in biological research with the Illinois Department of Agriculture, Conservation, and Public Health; the University of Illinois, Southern Illinois University, and other educational institutions; various research branches of the federal government; and other agricultural, conservation, municipal, and business organizations throughout the state.
INHS cooperates in biological research with the Illinois Department of Agriculture, Conservation, and Public Health; the University of Illinois, Southern Illinois University, and other educational institutions; various research branches of t he federal government; and other agricultural, conservati on, municipal, and bus iness organizations throughout t he state.
INHS aquatic biologists were involved in a 4-year study of the Kankakee River and Horse Creek near Custer Park, Illinois. The purpose of the study is to obtain biological, physical, and chemical data which will be used to evaluate any effects of the construction and operation of the Braidwood Station and its associated cooling lake on the biota and water quality of the Kankakee River and Horse Creek. The station's cooling pond will use the Kankakee River as a source of water for both intake and discharge purposes.
INHS aquatic biologists were involved in a 4
1.4.5.10 NUS Corporation NUS Corporation is a consulting engineering, research, and testing firm specializing in environmental and energy systems engineering, systems analysis, design engineering, management consulting, and training programs related to these areas. NUS has provided advice and professional guidance to utility, industrial, and government clients throughout the United States and in a number of foreign countries.
-year study of the Kankakee River and Hor se Creek near Custer P ark, Illinois. The purpose of the s tudy is to obtain bi ological, physical, and chemical data which will be used to evaluate any effects of the construction and operation of the Braidwood Station and its associated cooling lake on the biota and water quality of the Kankakee River and Hor se Creek. The sta tion's cooling pond will  
As a technical consultant on the Braidwood project, NUS was involved in a study to determine the adequacy of the station's ultimate heat sink.
1.4.5.11 Eberline Instrument Corporation (EIC)
Eberline Instrument Corporation (EIC) has provided radiation measurement equipment, comprehensive radiation protection services, and analytical laboratory services to the nuclear industry since 1953.
As a technical consultant on the Byron/Braidwood projects EIC performed preoperational environmental radiological baseline studies on and around the site.
1.4.5.12 Meteorology Research, Inc. (MRI)
Meteorology Research, Inc. (MRI) is an environmental consulting firm which, since 1951, has provided meteorological and air


use the Kankakee River as a source of water for both intake and discharge purposes.  
B/B-UFSAR 1.4-6 REVISION 1 - DECEMBER 1989 quality instruments and services to all aspects of industry in the solution of weather-related problems. These range from environmental impact assessments of existing or proposed airports and other major developments to problems of warehousing and marketing seasonal consumer goods. Of particular interest is the influence the local topography has on temperatures and winds.
MRI provided meteorological data from 1973 through mid-1975 for Byron and Braidwood Stations by implementation of an onsite meteorological measurement program.
1.4.5.13 Illinois State Museum (ISM)
The Illinois State Museum conducts archaeological investigations throughout the state of Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing and excavating, and laboratory analyses of datifacts, pollen, and soils.
As a technical consultant on the Braidwood project, ISM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Braidwood Station and pipeline corridor property.
1.4.5.14 Equitable Environmental Health, Inc. (EEH)
Equitable Environmental Health, Inc. (EEH), successor to Environmental Analysts, Inc./Tabershaw-Cooper Associated, Inc.,
is a multidisciplinary firm that offers the consulting services of medical professionals, scientists, engineers, economists, and technical support personnel in all areas of environmental health and economics.
EEH staff biologists conducted a 2-year baseline study of the Byron Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Rock River in the vicinity of the site were determined and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation on the biotic communities of the site were predicted, and data were provided for a benefit-cost analysis of the project.
1.4.5.15 Espey, Huston & Associates, Inc. (EH & A)
Espey, Huston & Associates, Inc. (EH & A) is a consulting firm addressing the environmental problems associated with industrial and urban development. EH & A professionals cover a broad range of expertise including civil engineering, environmental engineering, mathematics and computer science, and all phases of aquatic, estuarine, and terrestrial ecology.  


1.4.5.10 NUS Corporation
B/B-UFSAR 1.4-7 REVISION 1 - DECEMBER 1989 As a technical consultant on the Byron project, EH & A conducted the construction phase terrestrial and aquatic monitoring programs.
1.4.5.16 University of Wisconsin-Milwaukee (UWM)
The University of Wisconsin-Milwaukee under Dr. Elizabeth Benchley of the Dept. of Anthropology, conducts archaeological investigations throughout Wisconsin and northern Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing, and lab analysis of datifacts, pollen, and soils.
As a technical consultant on the Byron project, UWM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Byron Station and pipeline corridor property. Also, UWM conducted archaeological investigations on the Byron transmission line right-of-ways.
1.4.5.17 Aero-Metric Engineering, Inc. (AME)
Aero-Metric Engineering, Inc., founded in 1969, is based in Sheboygan, Wisconsin. The staff was made up of over 50 technical photogrammetric personnel, many having professional engineer and/or survey registration. AME's capabilities allow for a complete range of precision photogrammetric services, including aerial photography, mapping, and multiple survey skills.
As a technical consultant on the Byron project, AME will be providing annual aerial infra-red photographs.
1.4.5.18 Iowa Institute of Hydraulic Research The Iowa Institute of Hydraulic Research, formally organized in 1931, is a Division of the University of Iowa's College of Engineering. The Institute staff exceeded 80 in number and was comprised of a professional staff with Ph.Ds in the areas of Civil Engineering, Mechanical Engineering, Physics, Mechanics and Hydraulics, and Aeronautical Engineering, with most of these personnel holding joint academic appointments in the College of Engineering's Division of Energy Engineering. The Institute of Hydraulic Research conducts programs of fundamental research and advanced design and analysis in the areas of environmental pollution, bioengineering, naval hydrodynamics, river mechanics, ice hydraulics, hydrology, water resources, hydraulic structures, fluid mechanics, advanced instrumentation and data-handling techniques for fluids research, and mathematical modeling of watersheds and hydrology.


NUS Corporation is a consulting engineer ing, research, and testing firm specializ ing in environment al and energ y systems engineering, systems analysis, design engine ering, management consulting, and training pro grams related to these areas. NUS has provided advice and professional g uidance to utility, industrial, and governme nt clients throughou t the United States and in a number of f oreign countries.  
B/B-UFSAR 1.4-8 REVISION 9 - DECEMBER 2002 As a technical consultant on the Braidwood project, the Institute conducted a thermal evaluation to determine the adequacy of the ultimate heat sink.
1.4.5.19 Babcock and Wilcox International (B&W)
B&W is located in Cambridge, Ontario, Canada. B&W has fabricated fossil-fueled boiler components for over 100 years and has fabricated nuclear system components since the late 1950's. B&W has supplied replacement steam generators for Byron Unit 1 and Braidwood Unit 1.
1.4.5.20 Framatome Technologies, Incorporated (FTI)
FTI is located in Lynchburg, Virginia and has been providing services to the electric power industry for over four decades.
FTI engineering services include the necessary expertise, experience, and NRC-approved computer codes and methodologies to support the transient analysis of the Unit steam generators.
1.4.5.21 Stone & Webster Engineers and Constructors, Inc. (S&W)
S&W is located in Boston, Massachusetts and has been providing services to the electric power industry for over 100 years. S&W has provided balance-of-plant design-engineering support services in support of the power uprate of the Byron and Braidwood units.  


As a technical consult ant on the Braidwo od project, NUS was involved in a study to determine the adequacy of the station's ultimate heat sink.  
B/B-UFSAR 1.4-9 REVISION 8 - DECEMBER 2000 TABLE 1.4-1 EXELON GENERATION COMPANY'S NUCLEAR POWER PLANTS IN SERVICE OR UNDER CONSTRUCTION UNIT NOMINAL GROSS1 RATING (MWe)
SCHEDULED COMMERCIAL SERVICE DATE Dresden 1 210 1960 Dresden 2 850 1972 Dresden 3 850 1972 Quad-Cities 1 850 1972 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 La Salle 1 1122 1978 La Salle 2 1122 1979 Byron 1 1175 1985 Byron 2 1175 1987 Braidwood 1 1175 1988 Braidwood 2 1175 1988 1Note that this is a gross rating, not a net rating.  


1.4.5.11 Eberline Instrum ent Corporation (EIC)
B/B-UFSAR 1.4-10 REVISION 1 - DECEMBER 1989 TABLE 1.4-2 NUCLEAR POWER PLANTS COMPLETED OR CURRENTLY UNDER DESIGN BY SARGENT & LUNDY UNIT NOMINAL GROSS2 RATING (MWe)
YEAR OF POWER OPERATION EBWR 5
1956 Elk River 22 1962 La Crosse 60 1967 SEFOR 20 (MWt) 1969 Dresden 2 850 1969 Dresden 3 850 1971 Quad-Cities 1 850 1971 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 Fort St. Vrain, Unit 1 330 1973 Enrico Fermi, Unit 2 1200 1988 La Salle County Station, Unit 1 1122 1979 La Salle County Station, Unit 2 1122 1980 Byron Station, Unit 1 1175 1985 Byron Station, Unit 2 1175 1987 Braidwood Station, Unit 1 1175 1988 Braidwood Station, Unit 2 1175 1988 Clinton Power Station, Unit 1 992 1981 Kaiseraugst 992 1982 2Note that this is a gross rating, not a net rating.


Eberline Instrument Corp oration (EIC) has pr ovided radiation measurement equipment, comprehensive rad iation protection services, and analytical labor atory services to the nuclear industry since 1953.
B/B-UFSAR 1.4-11 TABLE 1.4-3 WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTS PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Shippingport Duquesne Light Company; Energy Research & Development Administration Pennsylvania 1957 90 4
As a technical consult ant on the Byron/Braidwood projects EIC performed preoperational envir onmental radiolo gical baseline studies on and a round the site.
Yankee-Rowe Yankee Atomic Electric Company Massachusetts 1961 175 4
1.4.5.12 Meteorology Re search, Inc. (MRI)
Trio Vercellese (Enrico Fermi)
Meteorology Research, In
Ente Nazionale per L'Energia Elettrica (ENEL)
: c. (MRI) is an environ mental consulting firm which, since 19 51, has provided m eteorological and air
Italy 1965 260 4
Chooz (Ardennes)
Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA)
France 1967 305 4
San Onofre Unit 1 Southern California Edison Co.;
San Diego Gas and Electric Co.
California 1968 450 3
Haddam Neck (Connecticut Yankee)
Connecticut Yankee Atomic Power Company Connecticut 1968 575 4
Jose Cabrera-Zorita Union Electrica, S.A.
Spain 1969 153 1
Beznau Unit 1 Nordostschweizerische Krafwerke AG (NOK)
Switzerland 1969 350 2
Robert Emmett Ginna Rochester Gas and Electric Corporation New York 1970 490 2
Mihama Unit 1 The Kansai Electric Power Company, Inc.
Japan 1970 320 2
Point Beach Unit 1 Wisconsin Electric Power Co.;
Wisconsin Michigan Power Co.
Wisconsin 1970 497 2
H. B. Robinson Unit 2 Carolina Power and Light Co.
South Carolina 1971 707 3


B/B-UFSAR 1.4-6 REVISION 1 - DECEMBER 1989 quality instruments and services to all aspects of industry in the solution of weathe r-related problems.
B/B-UFSAR 1.4-12 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
These range from environmental impact a ssessments of existing or proposed airports and other major deve lopments to proble ms of warehousing and marketing seasonal con sumer goods. Of particu lar interest is the influence the local to pography has on temper atures and winds.
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beznau Unit 2 Nordostschweizerische Kraftwerke AG (NOK)
MRI provided meteorolo gical data from 1973 t hrough mid-1975 for Byron and Braidwood St ations by implementati on of an onsite meteorological measu rement program.
Switzerland 1972 350 2
1.4.5.13 Illinois Sta te Museum (ISM)
Point Beach Unit 2 Wisconsin Electric Power Co.;
Wisconsin Michigan Power Co.
Wisconsin 1972 497 2
Surry Unit 1 Virginia Electric and Power Co.
Virginia 1972 822 3
Turkey Point Unit 3 Florida Power and Light Co.
Florida 1972 745 3
Indian Point Unit 2 Consolidated Edison Company of New York, Inc.
New York 1973 873 4
Prairie Island Unit 1 Northern States Power Company Minnesota 1973 530 2
Turkey Point Unit 4 Florida Power and Light Co.
Florida 1973 745 3
Surry Unit 2 Virginia Electric and Power Co.
Virginia 1973 822 3
Zion Unit 1 Exelon Generation Company Illinois 1973 1050 4
Kewaunee Wisconsin Public Service Corp.;
Wisconsin Power and Light Co.;
Madison Gas and Electric Co.
Wisconsin 1974 560 2
Prairie Island Unit 2 Northern States Power Company Minnesota 1974 530 2
Takahama Unit 1 The Kansai Electric Power Company, Inc.
Japan 1974 781 3
Zion Unit 2 Exelon Generation Company Illinois 1974 1050 4


The Illinois State Mus eum conducts archaeolo gical investigations throughout the s tate of Illinois. As a member of the Illinois Archaeological Survey, t hey have the experti se and services to perform contract archaeo logical work. Their studies included a pedestrian reconnaissance surv ey, subsurface testing and excavating, and laborato ry analyses of datifac ts, pollen, and soils.
B/B-UFSAR 1.4-13 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
As a technical consult ant on the Braid wood project, ISM identified and made re commendations which Commonwealth Edison acted upon to aid in preserv ing the archaeolog ical sites on Braidwood Station and pi peline corridor property.  
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Doel Unit 1 Indivision Doel Belgium 1975 390 2
Doel Unit 2 Indivision Doel Belgium 1975 390 2
Donald C. Cook Unit 1 Indiana and Michigan Electric Company (AEP)
Michigan 1975 1060 4
Ringhals Unit 2 Statens Vattenfallsverk (SSPB)
Sweden 1975 822 3
Almaraz Unit 1 Unit Electrica, S.A.;
Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A.
Spain 1976 902 3
Beaver Valley Unit 1 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Company Pennsylvania 1976 852 3
Diablo Canyon Unit 1 Pacific Gas and Electric Co.
California 1976 1084 4
Indian Point Unit 3 Consolidated Edison Company of New York, Inc.
New York 1976 965 4
Lemoniz Unit 1 Iberduero, S.A.
Spain 1976 902 3
Salem Unit 1 Public Service Electric and Gas Company; Exelon Generation Company; Atlantic City Electric Co.;
Delmarva Power and Light Co.
New Jersey 1976 1090 4


1.4.5.14 Equitable Environmen tal Health, Inc. (EEH)
B/B-UFSAR 1.4-14 TABLE 1.4-3 (Cont'd)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Trojan Portland General Electric Co.;
Eugene Water and Electric Board; Pacific Power and Light Company Oregon 1976 1130 4
Almaraz Unit 2 Union Electrica, S.A.;
Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A.
Spain 1977 902 3
Asco Unit 1 Fuerzas Electricas de Cataluna, S.A. (FESCA)
Spain 1977 902 3
Diablo Canyon Unit 2 Pacific Gas and Electric Co.
California 1977 1106 4
Joseph M. Farley Unit 1 Alabama Power Company Alabama 1977 829 3
Ko-Ri Unit 1 Korea Electric Power Co., Ltd.
Korea 1977 564 2
North Anna Unit 1 Virginia Electric and Power Co.
Virginia 1977 898 3
North Anna Unit 2 Virginia Electric and Power Co.
Virginia 1977 898 3
Ohi Unit 1 The Kansai Electric Power Co., Inc.
Japan 1977 1122 4
Ohi Unit 2 The Kansai Electric Power Co., Inc.
Japan 1977 1122 4
Ringhals Unit 3 Statens Vattenfallsvert (SSPB)
Sweden 1977 900 3
Sequoyah Unit 1 Tennessee Valley Authority Tennessee 1977 1148 4
Angra dos Reis Unit 1 Furnas-Centrais Electricas, S.A.
Brazil 1978 626 2


Equitable Environmen tal Health, Inc. (
B/B-UFSAR 1.4-15 TABLE 1.4-3 (Cont'd)
EEH), successor to Environmental Analysts, Inc./Tabershaw-Cooper Associated, Inc.,
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Asco Unit 2 Fuerzas Electricas de Cataluna, S.A. (FESCA);
is a multidisciplinary firm that offers the consulting services of medical professionals, sc ientists, engineer s, economists, and technical support pe rsonnel in all areas of environmental health and e conomics.
Empresa Nacional Hidroelectrica del Ribagorzana, S.A. (ENHER);
Fuerzas Hidroelectricas del Segre, S.A.;
Hidroelectrica de Cataluna, S.A.
Spain 1978 902 3
Donald C. Cook Unit 2 Indiana and Michigan Electric Company (AEP)
Michigan 1978 1060 4
Lemoniz Unit 2 Iberduero, S.A.
Spain 1978 902 3
Sequoyah Unit 2 Tennessee Valley Authority Tennessee 1978 1148 4
Watts Bar Unit 1 Tennessee Valley Authority Tennessee 1978 1177 4
William B. McGuire Unit 1 Duke Power Company North Carolina 1978 1180 4
Joseph M. Farley Unit 2 Alabama Power Company Alabama 1979 829 3
Krsko Savske Elektrarne, Ljubljana, Slovenia, Elektroprivreda, Zagreb, Croatia Yugoslavia 1979 615 2
Ringhals Unit 4 Statens Vattenfallsvert (SSPD)
Sweden 1979 900 3


EEH staff biologists conducted a 2-year base line study of the Byron Station site. Dis tributions of phytopla nkton, zooplankton, periphyton, benthos, f ish, fish eggs and larvae, and water chemistry in the Rock River in t he vicinity of t he site were determined and q uantitative data on terrestr ial flora and fauna were collected. The impacts of plant constr uction and operation on the biotic communities of the site were predicted, and data were provided for a benefit-cost analysis of the project.
B/B-UFSAR 1.4-16 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
1.4.5.15 Espey, Huston & Asso ciates, Inc. (EH & A)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Salem Unit 2 Public Service Electric and Gas Company; Exelon Generation Company Atlantic City Electric Co.;
Delmarva Power and Light Co.
New Jersey 1979 1115 4
Virgil C. Summer South Carolina Electric and Gas Company South Carolina 1979 900 3
Watts Bar Unit 2 Tennessee Valley Authority Tennessee 1979 1177 4
William B. McGuire Unit 2 Duke Power Company North Carolina 1979 1180 4
Byron Unit 1 Exelon Generation Company Illinois 1981 1120 4
Comanche Peak Unit 1 Texas Utilities Generating Co.
Texas 1980 1150 4
Seabrook Unit 1 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1980 1200 4
South Texas Project Unit 1 Houston Lighting and Power Co.;
Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Texas 1980 1250 4


Espey, Huston & Associat es, Inc. (EH & A) is a consulting firm addressing the e nvironmental problems as sociated with industrial and urban development.
B/B-UFSAR 1.4-17 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
EH & A professionals c over a broad range of expertise includi ng civil engineeri ng, environmental engineering, mathematics and com puter science, and all phases of aquatic, estuarine, and terrestrial ecology.
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beaver Valley Unit 2 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Co.;
B/B-UFSAR 1.4-7 REVISION 1 - DECEMBER 1989 As a technical consultant on the Byron project, EH & A conducted the construction p hase terrestrial and aquatic monitoring programs.
Cleveland Electric Illuminating Company; Toledo Edison Company Pennsylvania 1981 852 3
1.4.5.16 University of Wi sconsin-Milwaukee (UWM)
Braidwood Unit 1 Exelon Generation Company Illinois 1981 1120 4
The University of Wi sconsin-Milwaukee un der Dr. Elizabeth Benchley of the Dept. of Anthrop ology, conducts archaeological investigations throughout Wisconsin and nort hern Illinois. As a member of the Illinois Archaeo logical Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestri an reconnaissance survey, subsurface t esting, and lab anal ysis of datifacts, pollen, and soils.
Callaway Unit 1 SNUPPS - Union Electric Co.
As a technical consult ant on the Byron p roject, UWM identified and made recommendations which C ommonwealth Edison acted upon to aid in preserving the arc haeological sites on Byron Station and pipeline corrido r property. Also, UWM conducted archaeological investiga tions on the Byron transmission line right-of-ways.
Missouri 1981 1150 4
1.4.5.17 Aero-Metric En gineering, Inc. (AME)
Catawba Unit 1 Duke Power Company South Carolina 1981 1153 4
Aero-Metric Engineering, Inc., f ounded in 1969, is based in Sheboygan, Wisconsin.
Jamesport Unit 1 Long Island Lighting Company New York 1981 1150 4
The staff was made up of over 50 technical photogrammetric person nel, many having professional engineer and/or survey registration. AM E's capabilities allow for a complete range of precision phot ogrammetric services, including aerial photography, mapping, and m ultiple survey skills.
Ko-Ri Unit 2 Korea Electric Power Co., Ltd.
As a technical consult ant on the Byron project, AME will be providing annual aerial infra-red photographs.
Korea 1981 605 2
1.4.5.18 Iowa Institute of Hydraulic Research The Iowa Institute of Hy draulic Research, form ally organized in 1931, is a Division of the University of Iowa's College of Engineering. The Institute staff exceeded 80 in number and was comprised of a profess ional staff with Ph.Ds in the areas of Civil Engineering, Mec hanical Engineering, P hysics, Mechanics and Hydraulics, and Aero nautical Engineering, with most of these personnel holding joint academic a ppointments in the College of Engineering
NORCO Puerto Rico Water Resources Authority Puerto Rico 583 2
's Division of Ene rgy Engineering. The Institute of Hydraul ic Research conducts programs of fundamental research and advanced design and analysis in the areas of environmental pollution, bioeng ineering, naval hydrodynamics, river mechanics, ice hydraulics, hydrology, water resources, hydraulic struc tures, fluid m echanics, advanced instrumentation and dat a-handling techniques for fluids research, and m athematical modeling of watersheds and hydrology.
Braidwood Unit 2 Exelon Generation Company Illinois 1982 1120 4
Byron Unit 2 Exelon Generation Company Illinois 1982 1120 4
Catawba Unit 2 Duke Power Company South Carolina 1982 1153 4
Comanche Peak Unit 2 Texas Utilities Generating Co.
Texas 1982 1150 4
Marble Hill Unit 1 Public Service Company of Indiana, Inc.;
Northern Indiana Public Service Company Indiana 1982 1150 4


B/B-UFSAR 1.4-8 REVISION 9 - DECEMBER 2002 As a technical consult ant on the Braid wood project, the Institute conducted a thermal evaluation to determine the adequacy of the ulti mate heat sink.
B/B-UFSAR 1.4-18 TABLE 1.4-3 (Cont'd)
1.4.5.19 Babcock and Wilc ox International (B&W)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Millstone Unit 3 Northeast Nuclear Energy Co.
B&W is located in Cambri dge, Ontario, Canada.
Connecticut 1982 1156 4
B&W has fabricated fossil-fueled boiler c omponents for over 100 years and has fabricated nuclear system compon ents since the l ate 1950's. B&W has supplied replaceme nt steam generators fo r Byron Unit 1 and Braidwood Unit 1.
Seabrook Unit 3 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1982 1200 4
South Texas Project Unit 2 Houston Lighting and Power Co.;
Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Texas 1982 1250 4
Taiwan Unit 5 Taiwan Power Company Taiwan 1982 950 3
Wolf Creek Unit 1 SNUPPS - Kansas Gas and Electric Company; Kansas City Power and Light Company Kansas 1982 1150 4
Alvin W. Vogtle Unit 1 Georgia Power Company Georgia 1983 1113 4
Callaway Unit 2 SNUPPS - Union Electric Company Missouri 1983 1150 4
NEP-1 New England Power Company 1983 1150 4
Fort Calhoun Unit 2 Omaha Public Power District; Nebraska Public Power District Nebraska 1983 1150 4
Jamesport Unit 2 Long Island Lighting Company New York 1983 1150 4
Sears Island Central Maine Power Company Maine 1200 4


1.4.5.20 Framatome Technologi es, Incorporated (FTI)
B/B-UFSAR 1.4-19 TABLE 1.4-3 (Cont'd)
 
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Taiwan Unit 6 Taiwan Power Company Taiwan 1983 950 3
FTI is located in Lync hburg, Virginia and has been providing services to the electric power i ndustry for over four decades.
Alvin W. Vogtle Unit 2 Georgia Power Company Georgia 1984 1113 4
FTI engineering services include the necessary expertise, experience, and NRC-appr oved computer codes and methodologies to support the transient analysis of the Un it steam generators.
Marble Hill Unit 2 Public Service Company of Indiana, Inc.;
1.4.5.21 Stone & Webster Engineers and Construc tors, Inc. (S&W)
Northern Indiana Public Service Company Indiana 1984 1150 4
S&W is located in Bost on, Massachusetts and has been providing services to the electric power industry for over 100 years. S&W has provided balance-o f-plant design-engineeri ng support services in support of the po wer uprate of the Byron and Braidwood units.
Shearon Harris Unit 1 Carolina Power and Light Co.
B/B-UFSAR  
North Carolina 1984 900 3
 
Sterling SNUPPS - Rochester Gas and Electric Corporation; Central Hudson Gas and Electric Corporation; Niagara Mohawk Power Corporation; Orange and Rockland Utilities, Inc.
1.4-9 REVISION 8
New York 1984 1150 4
- DECEMBER 2000 TABLE 1.4-1 EXELON GENERATION COMPAN Y'S NUCLEAR POWER PLANTS IN SERVICE OR UNDER CONSTRUCTION UNIT NOMINAL GROSS 1 RATING (MWe)
Atlantic Unit 1 (O.P.S.)
SCHEDULED COMMERCIAL SERVICE DATE    Dresden 1 210 1960 Dresden 2 850 1972 Dresden 3 850 1972 Quad-Cities 1 850 1972 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 La Salle 1 1122 1978 La Salle 2 1122 1979 Byron 1 1175 1985 Byron 2 1175 1987 Braidwood 1 1175 1988 Braidwood 2 1175 1988 1Note that this is a gross rating, not a net rating.
Public Service Electric and Gas Company; Atlantic City Electric Co.;
B/B-UFSAR 1.4-10 REVISION 1 - DECEMBER 1989 TABLE 1.4-2 NUCLEAR POWER PLANTS COMPLETED OR CURRENTLY UNDER DESIGN BY SARGENT & LUNDY UNIT NOMINAL GROSS 2 RATING (MWe)
Jersey Central Power and Light Company New Jersey 1985 1150 4
YEAR OF POWER OPERATION    EBWR 5 1956 Elk River 22 1962 La Crosse 60 1967 SEFOR 20 (MWt) 1969 Dresden 2 850 1969 Dresden 3 850 1971 Quad-Cities 1 850 1971 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 Fort St. Vrain, Unit 1 330 1973 Enrico Fermi, Unit 2 1200 1988 La Salle County Station, Unit 1 1122 1979 La Salle County Station, Unit 2 1122 1980 Byron Station, Unit 1 1175 1985 Byron Station, Unit 2 1175 1987 Braidwood Statio n, Unit 1 1175 1988 Braidwood Statio n, Unit 2 1175 1988 Clinton Power Station, Unit 1 992 1981 Kaiseraugst 992 1982 2Note that this is a gross rating, not a net rating.
NEP-2 New England Power Company 1985 1150 4
B/B-UFSAR 1.4-11 TABLE 1.4-3 WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTS PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS        Shippingport Duquesne Light Company; Energy Research & Development Administration Pennsylvania 1957 90 4 Yankee-Rowe Yankee Atomic Electric Company Massachusetts 1961 175 4 Trio Vercellese (Enrico Fermi)
South Dade Unit 1 Florida Power and Light Co.
Ente Nazionale per L'Energia Elettrica (ENEL) Italy 1965 260 4        Chooz (Ardennes)
Florida 1985 1150 4
Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA) France 1967 305 4        San Onofre Unit 1 Southern California Edison Co.; San Diego Gas and Electric Co. California 1968 450 3 Haddam Neck (Connecticut Yankee)
Sundesert Unit 1 San Diego Gas and Electric Co.
Connecticut Yankee Atomic Power Company Connecticut 1968 575 4 Jose Cabrera-Zorita  Union Electrica, S.A. Spain 1969 153 1 Beznau Unit 1 Nordostschweizerische Krafwerke AG (NOK) Switzerland 1969 350 2 Robert Emmett Ginna  Rochester Gas and Electric Corporation New York 1970 490 2 Mihama Unit 1 The Kansai Electric Power Company, Inc. Japan 1970 320 2        Point Beach Unit 1 Wisconsin Electric Power Co.; Wisconsin Michigan Power Co. Wisconsin 1970 497 2 H. B. Robinson Unit 2  Carolina Power and Light Co. South Carolina 1971 707 3 B/B-UFSAR 1.4-12 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
California 1985 950 3  
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS       Beznau Unit 2 Nordostschweizerische Kraftwerke AG (NOK) Switzerland 1972 350 2 Point Beach Unit 2 Wisconsin Electric Power Co.; Wisconsin Michigan Power Co. Wisconsin 1972 497 2 Surry Unit 1 Virginia Electric and Power Co. Virginia 1972 822 3 Turkey Point Unit 3  Florida Power and Light Co. Florida 1972 745 3 Indian Point Unit 2  Consolidated Edison Company of New York, Inc. New York 1973 873 4 Prairie Island Unit 1  Northern States Power Company Minnesota 1973 530 2 Turkey Point Unit 4  Florida Power and Light Co. Florida 1973 745 3 Surry Unit 2 Virginia Electric and Power Co. Virginia 1973 822 3 Zion Unit 1 Exelon Generation Company Illinois 1973 1050 4 Kewaunee Wisconsin Public Service Corp.; Wisconsin Power and Light Co.;
Madison Gas and Electric Co. Wisconsin 1974 560 2 Prairie Island Unit 2  Northern States Power Company Minnesota 1974 530 2 Takahama Unit 1 The Kansai Electric Power Company, Inc. Japan 1974 781 3        Zion Unit 2 Exelon Generation Company Illinois 1974 1050 4
 
B/B-UFSAR 1.4-13 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Doel Unit 1 Indivision Doel Belgium 1975 390 2 Doel Unit 2 Indivision Doel Belgium 1975 390 2 Donald C. Cook Unit 1  Indiana and Michigan Electric Company (AEP) Michigan 1975 1060 4        Ringhals Unit 2 Statens Vattenfallsverk (SSPB) Sweden 1975 822 3 Almaraz Unit 1 Unit Electrica, S.A.; Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A. Spain 1976 902 3        Beaver Valley Unit 1  Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Company Pennsylvania 1976 852 3 Diablo Canyon Unit 1  Pacific Gas and Electric Co. California 1976 1084 4 Indian Point Unit 3  Consolidated Edison Company of New York, Inc. New York 1976 965 4 Lemoniz Unit 1 Iberduero, S.A. Spain 1976 902 3 Salem Unit 1 Public Service Electric and Gas Company; Exelon Generation Company; Atlantic City Electric Co.;
Delmarva Power and Light Co. New Jersey 1976 1090 4 B/B-UFSAR 1.4-14 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Trojan  Portland General Electric Co.; Eugene Water and Electric Board; Pacific Power and Light Company Oregon 1976 1130 4        Almaraz Unit 2 Union Electrica, S.A.; Compania Sevillana de Electricidad, S.A.;
Hidroelectrica Espanola, S.A. Spain 1977 902 3        Asco Unit 1 Fuerzas Electricas de Cataluna, S.A. (FESCA) Spain 1977 902 3        Diablo Canyon Unit 2  Pacific Gas and Electric Co. California 1977 1106 4 Joseph M. Farley Unit 1  Alabama Power Company Alabama 1977 829 3 Ko-Ri Unit 1 Korea Electric Power Co., Ltd. Korea 1977 564 2 North Anna Unit 1 Virginia Electric and Power Co. Virginia  1977 898 3 North Anna Unit 2 Virginia Electric and Power Co. Virginia 1977 898 3 Ohi Unit 1 The Kansai Electric Power Co., Inc. Japan 1977 1122 4 Ohi Unit 2 The Kansai Electric Power Co., Inc. Japan 1977 1122 4 Ringhals Unit 3 Statens Vattenfallsvert (SSPB) Sweden 1977 900 3 Sequoyah Unit 1 Tennessee Valley Authority Tennessee 1977 1148 4 Angra dos Reis Unit 1  Furnas-Centrais Electricas, S.A. Brazil 1978 626 2
 
B/B-UFSAR 1.4-15 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Asco Unit 2 Fuerzas Electricas de  Cataluna, S.A. (FESCA);
Empresa Nacional Hidroelectrica del Ribagorzana, S.A. (ENHER); Fuerzas Hidroelectricas del  Segre, S.A.;
Hidroelectrica de Cataluna, S.A. Spain 1978 902 3        Donald C. Cook Unit 2  Indiana and Michigan Electric Company (AEP) Michigan 1978 1060 4        Lemoniz Unit 2 Iberduero, S.A. Spain 1978 902 3 Sequoyah Unit  2 Tennessee Valley Authority Tennessee 1978 1148 4 Watts Bar Unit 1 Tennessee Valley Authority Tennessee 1978 1177 4 William B. McGuire Unit 1  Duke Power Company North Carolina 1978 1180 4 Joseph M. Farley Unit 2  Alabama Power Company Alabama 1979 829 3 Krsko  Savske Elektrarne, Ljubljana, Slovenia, Elektroprivreda, Zagreb, Croatia Yugoslavia 1979 615 2 Ringhals Unit 4 Statens Vattenfallsvert (SSPD) Sweden 1979 900 3
 
B/B-UFSAR 1.4-16 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Salem Unit 2 Public Service Electric and  Gas Company; Exelon Generation Company Atlantic City Electric Co.; Delmarva Power and Light Co. New Jersey 1979 1115 4 Virgil C. Summer South Carolina Electric and  Gas Company South Carolina 1979 900 3 Watts Bar Unit 2 Tennessee Valley Authority Tennessee 1979 1177 4 William B. McGuire Unit 2  Duke Power Company North Carolina 1979 1180 4 Byron Unit 1 Exelon Generation Company Illinois 1981 1120 4 Comanche Peak Unit 1  Texas Utilities Generating Co. Texas 1980 1150 4 Seabrook Unit 1 Public Service Company of    New Hampshire; United Illuminating Company New Hampshire 1980 1200 4 South Texas Project Unit 1  Houston Lighting and Power Co.;
Central Power and Light Co.;
City Public Service of San Antonio; City of Austin, Texas Texas 1980 1250 4 B/B-UFSAR 1.4-17 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Beaver Valley Unit 2  Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Co.;
Cleveland Electric  Illuminating Company; Toledo Edison Company Pennsylvania 1981 852 3 Braidwood Unit 1 Exelon Generation Company Illinois 1981 1120 4 Callaway Unit 1 SNUPPS - Union Electric Co. Missouri 1981 1150 4 Catawba Unit 1 Duke Power Company South Carolina 1981 1153 4 Jamesport Unit 1 Long Island Lighting Company New York 1981 1150 4 Ko-Ri Unit 2 Korea Electric Power Co., Ltd. Korea 1981 605 2 NORCO  Puerto Rico Water Resources Authority Puerto Rico
- 583 2        Braidwood Unit 2 Exelon Generation Company Illinois 1982 1120 4 Byron Unit 2 Exelon Generation Company Illinois 1982 1120 4 Catawba Unit 2 Duke Power Company South Carolina 1982 1153 4 Comanche Peak Unit 2  Texas Utilities Generating Co. Texas 1982 1150 4 Marble Hill Unit 1  Public Service Company of  Indiana, Inc.;
Northern Indiana Public Service Company Indiana 1982 1150 4 B/B-UFSAR 1.4-18 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Millstone Unit 3 Northeast Nuclear Energy Co. Connecticut 1982 1156 4 Seabrook Unit 3 Public Service Company of  New Hampshire; United Illuminating Company New Hampshire 1982 1200 4 South Texas Project Unit 2  Houston Lighting and Power Co.; Central Power and Light Co.; City Public Service of  San Antonio; City of Austin, Texas Texas 1982 1250 4        Taiwan Unit 5 Taiwan Power Company Taiwan 1982 950 3 Wolf Creek Unit 1 SNUPPS - Kansas Gas and  Electric Company; Kansas City Power and Light Company Kansas 1982 1150 4        Alvin W. Vogtle Unit 1  Georgia Power Company Georgia 1983 1113 4 Callaway Unit 2 SNUPPS - Union Electric Company Missouri 1983 1150 4 NEP-1  New England Power Company
- 1983 1150 4 Fort Calhoun Unit 2  Omaha Public Power District; Nebraska Public Power District Nebraska 1983 1150 4        Jamesport Unit 2 Long Island Lighting Company New York 1983 1150 4 Sears Island Central Maine Power Company Maine    - 1200 4 B/B-UFSAR 1.4-19 TABLE 1.4-3 (Cont'd)
PLANT  OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS        Taiwan Unit 6 Taiwan Power Company Taiwan 1983 950 3 Alvin W. Vogtle Unit 2 Georgia Power Company Georgia 1984 1113 4 Marble Hill Unit 2 Public Service Company   of Indiana, Inc.;
Northern Indiana Public   Service Company Indiana 1984 1150 4       Shearon Harris Unit 1 Carolina Power and Light Co. North Carolina 1984 900 3 Sterling SNUPPS - Rochester Gas and   Electric Corporation; Central Hudson Gas and Electric Corporation; Niagara Mohawk Power Corporation; Orange and Rockland Utilities, Inc. New York 1984 1150 4 Atlantic Unit 1 (O.P.S.)
Public Service Electric and   Gas Company; Atlantic City Electric Co.;
Jersey Central Power and   Light Company New Jersey 1985 1150 4 NEP-2 New England Power Company  
- 1985 1150 4 South Dade Unit 1 Florida Power and Light Co. Florida 1985 1150 4 Sundesert Unit 1 San Diego Gas and Electric Co. California 1985 950 3  


B/B-UFSAR 1.4-20 TABLE 1.4-3 (Cont'd)
B/B-UFSAR 1.4-20 TABLE 1.4-3 (Cont'd)
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBEROF LOOPS       Tyrone Unit 1 SNUPPS - Northern States   Power Company Wisconsin 1985 1150 4 Shearon Harris Unit 2 Carolina Power and Light Co. North Carolina 1986 900 3 South Dade Unit 2 Florida Power and Light Co. Florida 1986 1150 4 Atlantic No. (O.P.S.) Public Service Electric and   Gas Company; Atlantic City Electric Co.;
PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Tyrone Unit 1 SNUPPS - Northern States Power Company Wisconsin 1985 1150 4
Jersey Central Power and Light Company New Jersey 1987 1150 4 Shearon Harris Unit 4 Carolina Power and Light Co. North Carolina 1988 900 3 Sundesert Unit 2 San Diego Gas and Electric Co. California 1988 950 3 Sayago Unit 1 Iberduero, S.A. Spain 1980's 1000 3 Sayago Unit 4 Iberduero, S.A. Spain 1980's 1000 3 Shearon Harris Unit 3 Carolina Power and Light Co. North Carolina 1990 900 3 Unassigned Unit 1 (O.P.S.)
Shearon Harris Unit 2 Carolina Power and Light Co.
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1990 1150 4 Unassigned Unit 2 (O.P.S.)
North Carolina 1986 900 3
South Dade Unit 2 Florida Power and Light Co.
Florida 1986 1150 4
Atlantic No. (O.P.S.)
Public Service Electric and Gas Company; Atlantic City Electric Co.;
Jersey Central Power and Light Company New Jersey 1987 1150 4
Shearon Harris Unit 4 Carolina Power and Light Co.
North Carolina 1988 900 3
Sundesert Unit 2 San Diego Gas and Electric Co.
California 1988 950 3
Sayago Unit 1 Iberduero, S.A.
Spain 1980's 1000 3
Sayago Unit 4 Iberduero, S.A.
Spain 1980's 1000 3
Shearon Harris Unit 3 Carolina Power and Light Co.
North Carolina 1990 900 3
Unassigned Unit 1 (O.P.S.)
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1990 1150 4
Unassigned Unit 2 (O.P.S.)
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1992 1150 4  
Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1992 1150 4  


B/B-UFSAR 1.5-1 1.5 REQUIREMENTS FOR FURTH ER TECHNICAL INFORMATION The design of the Byron/Braidwood units is bas ed upon proven concepts which have been develop ed and success fully applied to the design of pressurized water reactor system
B/B-UFSAR 1.5-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION The design of the Byron/Braidwood units is based upon proven concepts which have been developed and successfully applied to the design of pressurized water reactor systems. There are currently no areas of research and development which are required for operation of this plant.
: s. There are currently no areas of resear ch and developme nt which are required for operati on of this plant.
At the time of issuance of construction permits for the Byron/
At the time of issuance of construction permits for the Byron/
Braidwood units, the Prelimina ry Safety Analys is Report (PSAR) and the standard design report which it referenced, RESAR-3, identified certain res earch and development pr ograms which were incomplete. These p rograms, which have been successfully completed, have provid ed technical informati on which has been used either to demonst rate the safety of design, more sharply define margins of conser vatism, or lead to design improvements.
Braidwood units, the Preliminary Safety Analysis Report (PSAR) and the standard design report which it referenced, RESAR-3, identified certain research and development programs which were incomplete. These programs, which have been successfully completed, have provided technical information which has been used either to demonstrate the safety of design, more sharply define margins of conservatism, or lead to design improvements.
Reference 1 presents descriptions of those safety-related research and development programs which have been carried out for, by, or in conjunc tion with Westinghouse Nuclear Energy Systems, and which are applicable to Wes tinghouse pressurized water reactors. The discussion which follows documents the completion of the construction permit stage rese arch programs.
Reference 1 presents descriptions of those safety-related research and development programs which have been carried out for, by, or in conjunction with Westinghouse Nuclear Energy Systems, and which are applicable to Westinghouse pressurized water reactors. The discussion which follows documents the completion of the construction permit stage research programs.
1.5.1 Programs Required for Plant Operation Two programs were iden tified as required for plant design and operation in the PSAR:  
1.5.1 Programs Required for Plant Operation Two programs were identified as required for plant design and operation in the PSAR:
: a. core stability evaluation and  
: a. core stability evaluation and
: b. fuel rod burst program.  
: b. fuel rod burst program.
 
Both programs are complete. The fuel rod burst program was completed at the time of the PSAR. The core stability evaluation program was not. A discussion of the core stability evaluation program follows.
Both programs are co mplete. The fuel rod burst program was completed at the time of the PSA R. The core sta bility evaluation program was not. A disc ussion of the core s tability evaluation program follows.
1.5.1.1 Core Stability Evaluation The program to establish means for the detection and control of potential xenon oscillations and for the shaping of the axial power distribution for improved core performance has been satisfactorily completed. See item 1, Reference 2, for a further discussion of the tests and results.
1.5.1.1 Core Sta bility Evaluation
1.5.2 Other Programs Not Required for Plant Operation The following programs were not complete at the time of the PSAR but are now satisfactorily complete.  
 
The program to establish means for the detection and control of potential xenon oscill ations and for the shaping of the axial power distribution for improved core per formance has been satisfactorily completed.
See item 1, Referen ce 2, for a further discussion of the tests and results.
1.5.2 Other Programs N ot Required for Plant Operation
 
The following programs w ere not complete at the time of the PSAR but are now satisfac torily complete.
 
B/B-UFSAR 1.5-2 1.5.2.1 Fuel Development Program for Operation at High Power Densities The program to demonst rate the satisfact ory operation of fuel at high burnup and p ower densities has been satisfactorily completed. See item 8, Reference 2, for a further discussion of the program a nd its results.
 
1.5.2.2 Blowdown Forces Program Westinghouse has completed B LODWN-2, an improved digital computer program for the calcula tion of local fluid pressures, flows and density transients in the primary coolant systems during a LOCA.  


BLODWN-2 is used to evaluate the effects of blowdown forces in this application. Refer to it em 15 in Reference 4 for a further discussion of the tests and results.
B/B-UFSAR 1.5-2 1.5.2.1 Fuel Development Program for Operation at High Power Densities The program to demonstrate the satisfactory operation of fuel at high burnup and power densities has been satisfactorily completed. See item 8, Reference 2, for a further discussion of the program and its results.
1.5.2.3 Blowdown Heat Transfer Testing (Form erly Titled Delayed Departure From N ucleate Boiling)
1.5.2.2 Blowdown Forces Program Westinghouse has completed BLODWN-2, an improved digital computer program for the calculation of local fluid pressures, flows and density transients in the primary coolant systems during a LOCA.
The NRC Acceptance Cri teria for Emergency Core Cooling Systems for Light-Water Powe r Reactors was issued in Section 50.46 of 10 CFR 50 on December 28, 19
BLODWN-2 is used to evaluate the effects of blowdown forces in this application. Refer to item 15 in Reference 4 for a further discussion of the tests and results.
: 73. It defines the basis and conservative assumptio ns to be used in the evaluation of the performance of emergency core cooling systems (ECCS).
1.5.2.3 Blowdown Heat Transfer Testing (Formerly Titled Delayed Departure From Nucleate Boiling)
Westinghouse believes that some of the conservatism of the criteria is associated w ith the manner in wh ich transient DNB phenomena are treated in the e valuation models.
The NRC Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Power Reactors was issued in Section 50.46 of 10 CFR 50 on December 28, 1973. It defines the basis and conservative assumptions to be used in the evaluation of the performance of emergency core cooling systems (ECCS).
Transient critical heat flux data presented at the 1972 specialists meeting of the Committee on Reactor Safe ty Technology (C REST) indicated that the time to DNB can be dela yed under transi ent conditions.
Westinghouse believes that some of the conservatism of the criteria is associated with the manner in which transient DNB phenomena are treated in the evaluation models. Transient critical heat flux data presented at the 1972 specialists meeting of the Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be delayed under transient conditions.
To demonstrate the c onservatism of the E CCS evaluation models, Westinghouse initiated a program to experiment ally simulate the blowdown phase of a LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blow down Heat Transfer Program, which was start ed early in 1976. T esting was completed in 1979. A DNB corr elation developed by Wes tinghouse from these test results is used in the ECCS analyses for Byron/Braidwood.
To demonstrate the conservatism of the ECCS evaluation models, Westinghouse initiated a program to experimentally simulate the blowdown phase of a LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blowdown Heat Transfer Program, which was started early in 1976. Testing was completed in 1979. A DNB correlation developed by Westinghouse from these test results is used in the ECCS analyses for Byron/Braidwood.
Objective The objective of the blo wdown heat transfer te st was to determine the time that DNB occurs under LOCA conditions. Thi s information was used to confirm a new Westinghou se transient DNB correlation.
Objective The objective of the blowdown heat transfer test was to determine the time that DNB occurs under LOCA conditions. This information was used to confirm a new Westinghouse transient DNB correlation.
The steady-state DNB data obtain ed from 15x15 and 17x17 test programs was used to assure th at the geometrical differences between the two fuel a rrays is correctly treat ed in the transient correlations.  
The steady-state DNB data obtained from 15x15 and 17x17 test programs was used to assure that the geometrical differences between the two fuel arrays is correctly treated in the transient correlations.  


B/B-UFSAR 1.5-3 Program The program was divided into two phases.
B/B-UFSAR 1.5-3 Program The program was divided into two phases. The Phase I tests started from steady-state conditions, with sufficient power to maintain nucleate boiling throughout the bundle, and progressed through controlled ramps of decreasing test section pressure or flow initiated DNB. By applying a series of controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pressures relevant to a PWR blowdown.
The Phase I tests started from steady-st ate conditions, with sufficient power to maintain nucleate boiling throughout the bundle, and progressed  
Phase I provided separate-effects data for heat transfer correlation development.
Typical parameters used for Phase I testing are shown in Table 1.5-1.
Phase II simulated PWR behavior during a LOCA to permit definition of the time delay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after establishment of typical steady-state operating conditions. The fluid transient was then initiated, and the rod power decay was programmed in such a manner as to simulate the actual heat input of fuel rods. The test was terminated when the heater rod temperatures reached a predetermined limit.
Typical parameters used for Phase II testing are shown in Table 1.5-2.
Test Description The experimental program was conducted in the J-Loop at the Westinghouse Forest Hills Facility with a full length 5x5 rod bundle simulating a section of a 15x15 fuel assembly to determine DNB occurrence under LOCA conditions.
The heater rod bundles used in this program were internally-heated rods, capable of a maximum linear power of 18.8 kW/ft, with a total power of 135 kW (for extended periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribution. Each rod was adequately instrumented with a total of 12 clad thermocouples.
Results The experiments in the DNB facility resulted in cladding temperature and fluid properties measured as a function of time throughout the blowdown range from 0 to 20 seconds.
Facility modifications and installation of the initial test bundle were completed. A series of shakedown tests in the


through controlled ram ps of decreasing test se ction pressure or flow initiated DNB.
B/B-UFSAR 1.5-4 REVISION 1 - DECEMBER 1989 J-Loop were performed. These tests provided data for instrumentation calibration and check-out, and provided information regarding facility control and performance. Initial program tests were performed during the first half of 1975.
By applying a series of controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pres sures relevant to a PWR blowdown.  
Under the sponsorship of EPRI, testing was reinitiated during 1976 on the same test bundle. The testing was terminated in November 1976 and plans were made for a new test bundle and further testing during 1978-1979. These tests were completed in December of 1979.
1.5.3 References
: 1.
F. T. Eggleston, "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, October 1978.
: 2.
F. T. Eggleston, "Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8768.
Spring 1976 Edition.
: 3.
"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8458. Fall 1977 Edition.
: 4.
"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8004. Fall 1972 Edition.  


Phase I provided separate-effects data for h eat transfer correlation development.  
B/B-UFSAR 1.5-5 TABLE 1.5-1 BLOWDOWN HEAT TRANSFER PHASE I TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 1250 to 2250 psia Test section mass velocity 1.12 to 2.5x106 lb/hr-ft2 Core inlet temperature 550° F to 600° F Maximum heat flux 306,000 to 531,000 Btu/hr-ft2 TRANSIENT RAMP CONDITIONS Pressure decrease 0 to 350 psia/sec and subcooled depressurization from 2250 psia Flow decrease 0 to 100%/sec Inlet enthalpy constant


Typical parameters used for Phase I test ing are shown in Table 1.5-1.
B/B-UFSAR 1.5-6 TABLE 1.5-2 BLOWDOWN HEAT TRANSFER PHASE II TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 2250 psia Test section mass velocity 2.5x106 lb/hr-ft2 Inlet coolant temperature 545° F Maximum heat flux 531,000 Btu/hr-ft2 TRANSIENT CONDITIONS Simulated break Double-ended cold leg guillotine breaks
Phase II simulated P WR behavior during a LOCA to permit definition of the time d elay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after establishment of typical steady-stat e operating cond itions. The fluid transient was th en initiated, and the rod power decay was programmed in such a m anner as to simulate t he actual heat input of fuel rods. The test was terminated when the heater rod temperatures reached a predetermined limit.


Typical parameters u sed for Phase II testing are shown in Table 1.5-2.  
B/B-UFSAR 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCES Table 1.6-1 lists topical reports which provide information additional to that provided in this UFSAR and which have been filed separately with the Nuclear Regulatory Commission (NRC) in support of this and similar applications.
A legend to the review status code letters follows:
A
- NRC review complete; NRC acceptance letter issued.
AE
- NRC accepted as part of the Westinghouse Emergency Core Cooling System (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.
B
- Submitted to the NRC as background information; not undergoing formal NRC review.
O
- On file with NRC; older generation report with current validity; not actively under formal NRC review.
U
- Actively under formal NRC review.  


Test Description
B/B-UFSAR 1.6-2 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE REPORT REFERENCE SECTION(S)
REVIEW STATUS "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," WCAP-2048, July 1962 4.3 0
"Single Phase Local Boiling and Bulk Boiling Pressure Drop Correlations," WCAP-2850 (Proprietary), April 1966 and WCAP-7916 (Non-Proprietary), June 1972 4.4 0
"In-Pile Measurement of UO2 Thermal Conductivity," WCAP-2923, 1966 4.4 0
"Hydraulic Tests of the San Onofre Reactor Model," WCAP-3269-8, June 1964 4.4 0
"LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM - 7094,"
WCAP-3269-26, September 1963 4.3, 4.4 15.0, 15.4 0
"Saxton Core II Fuel Performance Evaluation,"
WCAP-3385-56, Part II, "Evaluation of Mass Spectrometric and Radiochemical Analyses of Irradiated Saxton Plutonium Fuel," July 1970 4.3, 4.4 0
"Xenon-Induced Spatial Instabilities in Large PWRs," WCAP-3680-20, (EURAEC-1974)
March 1968 4.3 0
"Control Procedures for Xenon-Induced X-Y Instabilities in Large PWR's,"
WCAP-3680-21, (EURAEC-2111) February 1969 4.3 0
"Xenon-Induced Spatial Instabilities in Three-Dimensions," WCAP-3680-22, (EURAEC-2116) September 1969 4.3 0
"Pressurized Water Reactor pH - Reactivity Effect Final Report," WCAP-3698-8, (EURAEC-2074) October 1968 4.3 0
"PUO2 - UO2 Fueled Critical Experiments,"
WCAP-3726-I, July 1967 4.3 0


The experimental program was con ducted in the J-Loop at the Westinghouse Forest Hills Facili ty with a full length 5x5 rod bundle simulating a se ction of a 15x15 fuel as sembly to determine DNB occurrence under LOCA conditions.  
B/B-UFSAR 1.6-3 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Melting Point of Irradiated UO2,"
WCAP-6065, February 1965 4.2, 4.4 0
"Burnup Physics of Heterogeneous Reactor Lattices," WCAP-6069, June 1965 4.4 0
"LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS,"
WCAP-6073, April 1966 4.3 0
"Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium," WCAP-6086, August 1969 4.3 0
"Subchannel Thermal Analysis of Rod Bundle Cores," WCAP-7015, Revision 1, January 1969 4.4 0
"The PANDA Code," WCAP-7048 (Proprietary) and WCAP-7757 (Non-Proprietary), January 1975 4.3 A
"Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198-L (Proprietary), April 1969 and WCAP-7825 (Non-Proprietary), December 1971 4.3 0
"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7208 (Proprietary), September 1968 and WCAP-7811, (Non-Proprietary), December 1971 4.3 "The TURTLE 24.0 Diffusion Depletion Code,"
WCAP-7213 (Proprietary) and WCAP-7758 (Non-Proprietary), January 1975 4.3, 15.0 15.4 A
"Core Power Capability in Westinghouse PWRs,"
WCAP-7267-L (Proprietary), October 1969 and WCAP-7809 (Non-Proprietary), December 1971 4.3 "Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors,"
WCAP-7306, April 1969 15.4 "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7308, December 1971 4.3 A


The heater rod bundl es used in this prog ram were internally-heated rods, cap able of a maximum line ar power of 18.8 kW/ft, with a total power of 135 kW (for exte nded periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribut ion. Each rod was adequately instrumented with a total of 12 clad thermocouples.
B/B-UFSAR 1.6-4 TABLE 1.6-1 (Cont'd)
Results The experiments in t he DNB facility re sulted in cladding temperature and fluid pr operties measured as a function of time throughout the blowd own range from 0 to 20 seconds.
REPORT REFERENCE SECTION(S)
Facility modifications and installation of the initial test bundle were completed.
REVIEW STATUS "Application of the THINC Program to PWR Design," WCAP-7359-L (Proprietary), August 1969 and WCAP-7838 (Non-Proprietary),
A series of shak edown tests in the
January 1972 4.4 O
"Seismic Testing of Electrical and Control Equipment," WCAP-7397-L (Proprietary) and WCAP-7817 (Non-Proprietary), December 1971 3.10 O
"Seismic Testing of Electrical and Control Equipment (WCID Process Control Equipment),"
WCAP-7397-L, Supplement 1 (Proprietary) and WCAP-7817, Supplement 1 (Non-Proprietary),
December 1971 3.10 O
"Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," WCAP-7477-L (Proprietary), March 1970 and WCAP-7735 (Non-Proprietary), August 1971 5.2 A
"Radiological Consequences of a Fuel Handling Accident," WCAP-7518-L (Proprietary) and WCAP-7828 (Non-Proprietary), June 1970 15.7 O
"Seismic Vibration Testing with Sine Beats,"
WCAP-7558, October 1972 3.10 O
"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975 15.4 A
"Dynamic Fracture Toughness Properties of Heavy Section A533 Grade B Class 1 Steel Plate," WCAP-7623, December 1970 5.4 O
"Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L (Proprietary) and WCAP-7755 (Non-Proprietary), January 1975 4.4 A
"DNB Tests Results for New Mixing Vane Grids (R)," WCAP-7695-L (Proprietary) and WCAP-7958 (Non-Proprietary) and Addendum, January 1975 4.4 A


B/B-UFSAR 1.5-4 REVISION 1 - DECEMBER 1989 J-Loop were performed. These tests provided data for instrumentation calibr ation and check-ou t, and provided information regarding fa cility control a nd performance. Initial program tests were performed during the first half of 1975.
B/B-UFSAR 1.6-5 TABLE 1.6-1 (Cont'd)
Under the sponsorship of EPRI, t esting was rei nitiated during 1976 on the same test bu ndle. The testing was terminated in November 1976 and plans were m ade for a new test bundle and further testing during 1 978-1979. These tes ts were completed in December of 1979.  
REPORT REFERENCE SECTION(S)
REVIEW STATUS "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients,"
WCAP-7706, February 1973 4.6, 7.1 O
"Electric Hydrogen Recombiner for PWR Containments," WCAP-7709-L, Supplements 1 through 7 (Proprietary) and WCAP-7820, Supplements 1 through 7 (Non-Proprietary),
1971 through 1977 3.11, 6.2 A
"A Comprehensive Space-Time Dependent Analysis of Loss of Coolant (SATAN-IV Digital Code),"
WCAP-7750, August 1971 3.6 O
"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, October 1971 15.2 O
"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June 1972 5.2 O
"Behavior of Austenitic Stainless Steel in Post Hypothetical Loss of Coolant Accident Environment," WCAP-7798-L (Proprietary) and WCAP-7803 (Non-Proprietary), January 1972 6.1 O
"Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800, Revision 4-A, April 1975 4.2, 17 A
"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"
WCAP-7806, December 1971 4.3 B
"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7811, December 1971 4.3 O
"Seismic Testing of Electrical and Control Equipment (Low Seismic Plants)," WCAP-7817, Supplements 1-8, December 1971-March 1974 3.10 O


====1.5.3 References====
B/B-UFSAR 1.6-6 TABLE 1.6-1 (Cont'd)
: 1. F. T. Eggleston, "Safety-Rel ated Research and Development for Westinghouse Pressurized W ater Reactors, Program  
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Evaluation of Steam Generator Tube, Tubesheet and Divider Plate Under Combined LOCA Plus SSE Conditions," WCAP-7832, December 1973 5.4 A
"Inlet Orificing of Open PWR Cores,"
WCAP-7836, January 1972 4.4 B
"Neutron Shielding Pads," WCAP-7870, May 1972 3.9 A
"LOFTRAN Code Description," WCAP-7907, June 1972 5.2, 15.0 15.1, 15.2, 15.3, 15.4, 15.5, 15.6 A
"FACTRAN - A FORTRAN-IV Code for Thermal Transients in a UO2 Fuel Rod," WCAP-7908, June 1972 15.0, 15.2 15.3, 15.4 A
"MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop PWR System,"
WCAP-7909, June 1972 6.3 O
"Power Peaking Factors," WCAP-7912-L (Proprietary) and WCAP-7912 (Non-Proprietary), January 1975 and Supplement 4.3, 4.4 A
"Damping Values of Nuclear Power Plant Components," WCAP-7921, May 1974 lA, 3.7 A
"Basis for Heatup and Cooldown Limit Curves," WCAP-7924, April 1975 5.3 A
"Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid,"
WCAP-7941-L (Proprietary) and WCAP-7959 (Non-Proprietary), January 1975 4.4 A
"Fuel Assembly Safety Analysis for Combined Seismic and Loss of Coolant Accident, 15x15,"
WCAP-7950, July 1972 3.7 A
"THINC-IV An Improved Program for Thermal and Hydraulic Analysis of Rod Bundle Cores,"
WCAP-7956, June 1973 4.4 A


Summaries," WCAP-876 8, October 1978.  
B/B-UFSAR 1.6-7 REVISION 9 - DECEMBER 2002 TABLE 1.6-1 (Cont'd)
: 2. F. T. Eggleston, "Safety-Relat ed Research and Development for Westinghouse PWRs Program Summaries," WCAP-8768.
REPORT REFERENCE SECTION(S)
Spring 1976 Edition.
REVIEW STATUS "Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor," WCAP-7964, June 1971 4.3 O
: 3. "Safety-Related Rese arch and Development for Westinghouse PWRs Program Summaries," WCA P-8458. Fall 1977 Edition.  
"TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979 (Proprietary) and WCAP-8028 (Non-Proprietary), January 1975 15.0, 15.4 A
: 4. "Safety-Related Rese arch and Development for Westinghouse PWRs Program Summaries," WCA P-8004. Fall 1972 Edition.  
"WIT-6 Reactor Transient Analysis Computer Program Description," WCAP-7980, November 1972 15.0, 15.4 A
"Application of Modified Spacer Factor to "L" Grid Typical and Cold Wall Cell DNB,"
WCAP-7988 (Proprietary) and WCAP-8030 (Non-Proprietary), October 1972 4.4 A
"Application of the THINC-IV Program to PWR Design," WCAP-8054 (Proprietary) and WCAP-8195 (Non-Proprietary), October 1973 4.4 A
"Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop,"
WCAP-8082 (Proprietary) and WCAP-8172 (Non-Proprietary), January 1975 3.6 A
"Reactor Coolant Pump Integrity in LOCA,"
WCAP-8163, September 1973 lA, 5.4 O
"Calculational Model for Core Reflooding After a Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974 15.6 A
"Effect of Local Heat Flux Spikes on DNB in Non-Uniform Heated Rod Bundles," WCAP-8174 (Proprietary) and WCAP-8202, (Non-Proprietary), August 1973 4.4 A
"WFLASH, A FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWR," WCAP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision 1 (Non-Proprietary), July 1974 15.6 A


B/B-UFSAR 1.5-5 TABLE 1.5-1 BLOWDOWN HEAT TRANSFER P HASE I TEST PARAMETERS
B/B-UFSAR 1.6-8 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Fuel Densification Experimental Results and Model for Reactor Application,"
WCAP-8218 (Proprietary) and WCAP-8219 (Non-Proprietary), March 1975 4.1, 4.2, 4.3, 4.4 A
"Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236 (Proprietary), December 1973 and WCAP-8288 (Non-Proprietary), January 1974 and Addenda 3.7, 4.2 A
"Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March 1974 and WCAP-8288, Addendum 1 (Non-Proprietary), April 1974 3.7 A
"Documentation of Selected Westinghouse Structural Analysis Computer Codes,"
WCAP-8252, Revision 1, July 1977 3.6, 3.9 O
"Hydraulic Flow Test of the 17x17 Fuel Assembly," WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary), February 1974 4.2, 4.4 O
"Effect of 17x17 Fuel Assembly Geometry on DNB," WCAP-8296 (Westinghouse Proprietary) and WCAP-8927 (Non-Proprietary), February 1975 4.4 A
"The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing," WCAP-8298 (Proprietary) and WCAP-8299 (Non-Proprietary),
January 1975 4.4 A
"LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974 15.0, 15.6 AE SATAN-IV Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant,"
WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974 15.0, 15.6 AE


PARAMETERS NOMINAL VALUE  INITIAL STEADY-S TATE CONDITIONS Pressure 1250 to 2250 psia
B/B-UFSAR 1.6-9 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests,"
WCAP-8303 (Proprietary) and WCAP-8317 (Non-Proprietary), July 1975 3.9 A
"Control of Delta Ferrite in Austenitic Stainless Steel Weldments," WCAP-8324-A, June 1974 lA, 5.2 A
"Containment Pressure Analysis Code (COCO),"
WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974 15.6 AE "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974 4.3, 4.6, 15.1, 15.2, 15.4, 15.8 O
"Westinghouse ECCS Evaluation Model -
Summary," WCAP-8339, July 1974 6.2, 15.6 AE "Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974 15.6 AE "Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP -8341 (Proprietary) and WCAP-8342 (Non-Proprietary), July 1974 lA(N), 17 A
"Effects of Fuel Densification Power Spikes on Clad Thermal Transients," WCAP-8359, July 1974 4.3 AE "Westinghouse Nuclear Energy Systems Division Quality Assurance Plan," WCAP-8370, Revision 9A, September 1977 1A, 17 A
"Qualification of Westinghouse Seismic Testing Procedure for Electrical Equipment Tested Prior to May 1974," WCAP-8373, August 1974 3.10 O
"Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-Proprietary),
July 1974 4.2 A


Test section mass velocity 1.12 to 2.5x10 6 lb/hr-ft 2
B/B-UFSAR 1.6-10 TABLE 1.6-1 (Cont'd)
Core inlet temperature 550° F to 600
REPORT REFERENCE SECTION(S)
° F Maximum heat flux 306,000 to 531,000 Btu/hr-ft 2 
REVIEW STATUS "Power Distribution Control and Load Following Procedures," WCAP-8385 (Proprietary) and WCAP-8403 (Non-Proprietary), September 1974 4.3, 4.4 A
"An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," WCAP-8424, Revision 1, June 1975 15.3 O
"17x17 Drive Line Components Tests - Phase IB, II, III, D-Loop Drop and Deflection,"
WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December 1974 3.9, 15.0 A
"Analysis of Data from the Zion (Unit 1)
THINC Verification Test," WCAP-8453-A (Proprietary), May 1976 and WCAP-8454 (Non-Proprietary), January 1975 4.4 A
"Westinghouse ECCS Evaluation Model -
Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-Proprietary), April 1974 15.6 AE "Incore Power Distribution Determination in Westinghouse Pressurized Water Reactors,"
WCAP-8498, July 1975 4.3 O
"UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations," WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary), April 1975 3.9 A
"Critical Heat Flux Testing of 17x17 Fuel Assembly Geometry with 22 Inch Spacing,"
WCAP-8536 (Proprietary) and WCAP-8537 (Non-Proprietary), May 1975 4.4 A
"Westinghouse ECCS - Four Loop Plant (17x17)
Sensitivity Studies," WCAP-8565 (Proprietary) and WCAP-8566 (Non-Proprietary), July 1975 15.6 A


TRANSIENT RAMP CONDITIONS
B/B-UFSAR 1.6-11 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Improved Thermal Design Procedure,"
WCAP-8567-P, July 1975 (Proprietary) and WCAP-8568, July 1975 (Non-Proprietary) 4.4, 15.0 A
"Augmented Startup and Cycle 1 Physics Program Supplement 1," WCAP-8575, June 1976 (Proprietary) and WCAP-8576, June 1976 (Non-Proprietary) and Supplements.
4.3 O
"The Application of Preheat Temperatures After Welding Pressure Vessel Steels,"
WCAP-8577, February 1976 lA A
"Failure Mode and Effects Analysis (FMEA) of the Engineered Safeguard Features Actuation System," WCAP-8584 (Proprietary) and WCAP-8760 (Non-Proprietary), April 1976 4.6 O
"Environmental Qualification of Westinghouse NSSS Class lE Equipment," WCAP-8587, September 1975 lA, 3.1O, 3.11 A
"Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary), November 1975 15.6 A
"Experimental Verification of Wet Fuel Storage Criticality Analyses," WCAP-8682 (Proprietary) and WCAP-8683 (Non-Proprietary), December 1975 4.3 B
"Fuel Rod Bowing," WCAP-8691 (Proprietary) and WCAP-8692 (Non-Proprietary),
December 1975 4.2 O
"Delta Ferrite in Production Austenitic Stainless Steel Weldments," WCAP-8693, January 1976 lA, 5.2 B
"MULTIFLEX - A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708 (Proprietary) and WCAP-8709 (Non-Proprietary), February 1976 3.9 A


Pressure decrease 0 to 350 psia/sec and subcooled depressurization
B/B-UFSAR 1.6-12 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS Foster, J. P., et al., Westinghouse Improved Performance Analysis and Design Model (PAD 4.0), WCAP-15063-P-A, Revision 1 with Errata, July 2000.
4.2 A
"New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, July 1976 (Proprietary) and WCAP-8763, July 1976 (Non-Proprietary) 4.4 A
"Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, Revision 2, October 1978 1.5, 4.2, 4.3 B
"Verification of Neutron Pad and 17x17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant," WCAP-8780, May 1976 3.9 B
"Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations,"
WCAP-8785, October 1976 4.2 "Hybrid B4C Absorber Control Rod Evaluation Report," WCAP-8846, October 1977 4.2, 15.0 15.3 A
"Westinghouse ECCS - Four Loop Plant (17x17)
Sensitivity Studies with Upper Head Fluid Temperature at Thot," WCAP-8865, May 1977 15.6 A
"7300 Series Process Control System Noise Tests," WCAP-8892-A, April 1977 7.1 A
"Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8963 (Proprietary), November 1976 and WCAP-8964 (Non-Proprietary), August 1977 4.2 A
"Westinghouse Emergency Core Cooling System Small Break October 1975 Model," WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary),
April 1977 15.6 A


from 2250 psia
B/B-UFSAR 1.6-13 REVISION 1 - DECEMBER 1989 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Failure Mode and Effects Analysis of the Solid State Full Length Rod Control System,"
WCAP-8976, September 1977 4.6 O
"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"
WCAP-9000-L, Revision 1 (Proprietary), July 1969 and WCAP-7806 (Non-Proprietary), December 1971.
4.3 "Axial Power Distribution Monitoring Using Four-Section Ex-Core Detectors," WCAP-9105 (Proprietary) and WCAP-9106 (Non-Proprietary),
July 1977 4.3 A
"Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCAs During Operation with One Loop Out of Service for Plants Without Loop Isolation Valves,"
WCAP-9166 (Proprietary) and WCAP-9167 (Non-Proprietary), February 1978 15.6 O
"Westinghouse Emergency Core Cooling System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9150 (Non-Proprietary), September 1977 15.6 O
"Properties of Fuel and Core Component Materials," WCAP-9179 (Proprietary), September 1977 and WCAP-9224 (Non-Proprietary) 4.2 O
"Westinghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220 (Proprietary Version), WCAP-9221 (Non-Proprietary Version), February 1978 15.6 A
"Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly," WCAP-9401 (Proprietary) and WCAP-9402 (Non-Proprietary), March 1979 4.1, 4.2, 4.4 A
"PALADON - Westinghouse Nodal Computer Code,"
WCAP-9485 (Proprietary) and WCAP-9486 (Non-Proprietary) December 1978 4.3 A


Flow decrease 0 to 100%/sec
B/B-UFSAR 1.6-14 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)
REPORT REFERENCE SECTION(S)
REVIEW STATUS "Reference Core Report 17x17 Optimized Fuel Assembly," WCAP-9500 (Non Proprietary),
July 1979 4, 15 A
"RELAP5/MOD2-B&W - An Advanced Computer Code for Light Water Reactor LOCA and non-LOCA Transient Analysis" BAW-10164, Revision 3 (non-proprietary), October 1996 15 A
"CONTEMPT - Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident,", BAW-10095A, Revision 1, April 1978 6
O Beacon Core Monitoring and Operations Support System, WCAP-12472 (Proprietary Class 2),
August 1994 4.3, 4.4, 7.7 A
Relaxation of Constant Axial Offset Control, FQ Surveillance Technical Specification, WCAP-10216-P-A, Revision 1A (Proprietary Class 2), February 1994 4.3, 4.4 A
VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, WCAP-14565-P-A (Proprietary) / WCAP-15306-NP-A (Non-Proprietary), October 1999 4.4, 15.0 A
Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications, WCAP-14565-P-A Addendum 2-P-A (Proprietary) / WCAP-15306-NP-A Addendum 2-NP-A (Non-Proprietary), April 2008 4.4 A
SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, WCAP-10698-P-A (Proprietary) and WCAP-10750-A (Non-Proprietary), August 1987 15.0 A


Inlet enthalpy constant
B/B-UFSAR 1.7-1 REVISION 9 - DECEMBER 2002 1.7 DRAWINGS The drawings cited in each UFSAR Chapter 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. References on the figures contained in the UFSAR to ComEd, CECo, and Commonwealth Edison will be revised to reflect the change in facility ownership to Exelon Generation Company when other changes to that figure are needed.
 
1.7.1 Electrical, Instrumentation, and Control Drawings Subsection 1.7.1 of the FSAR provides a list of electrical, instrumentation, and control drawings that were provided to the NRC during the initial licensing phase.
B/B-UFSAR 1.5-6 TABLE 1.5-2 BLOWDOWN HEAT TRANSF ER PHASE II TEST PARAMETERS
1.7.2 Drawings for Independent Structural Review Subsection 1.7.2 of the FSAR provides a list of the structural, architectural, mechanical loading and electrical loading drawings that were provided to the NRC to enable them to perform the Project Structural Review and the Independent Structural Review during the licensing phase.  
 
PARAMETERS NOMINAL VALUE  INITIAL STEADY-S TATE CONDITIONS Pressure 2250 psia
 
Test section mass velocity 2.5x10 6 lb/hr-ft 2
Inlet coolant temperature 545° F Maximum heat flux 531,000 Btu/hr-ft 2 
 
TRANSIENT CONDITIONS Simulated break
 
Double-ended cold leg
 
guillotine breaks
 
B/B-UFSAR 1.6-1 1.6 MATERIAL INCORPO RATED BY REFERENCES Table 1.6-1 lists topical repo rts which prov ide information additional to that p rovided in this UFSAR and which have been filed separately with the Nuclear Regulatory Com mission (NRC) in support of this and si milar applications.
 
A legend to the revi ew status code l etters follows:
A - NRC review complete; NRC acceptance letter issued.
AE - NRC accepted as part of the Westinghouse Emergency Core Cooli ng System (ECCS) evaluation model only; does not constitute acceptance for any purpo se other than for ECCS analyses. B - Submitted to t he NRC as background information; not undergo ing formal NRC review. O - On file with NRC; ol der generation report with current validity; not actively under formal
 
NRC review. U - Actively under f ormal NRC review.
 
B/B-UFSAR 1.7-1 REVISION 9 - DECEMBER 2002  
 
===1.7 DRAWINGS===
The drawings cited in each UFSAR Chapter are included as "General References" only; i.
e., refer to the d rawings to obtain additional detail or to obtain background information. These drawings are not part of the UFS AR. They are controlled by the Controlled Documents Program. References on the figures contained in the UFSAR to ComE d, CECo, and Com monwealth Edison will be revised to r eflect the change in facility ownership to Exelon Generation Company when other c hanges to that figure are needed. 1.7.1 Electrical, Instrumentat ion, and Control Drawings
 
Subsection 1.7.1 of the FSAR provides a list of electrical, instrumentation, and control dra wings that were provided to the NRC during the initi al licensing phase.
1.7.2 Drawings for Indepen dent Structural Review
 
Subsection 1.7.2 of the FSAR provides a list of the structural, architectural, mechanical loading and electrical loading drawings that were provided to the NR C to enable them to perform the Project Structural Rev iew and the Indepe ndent Structural Review during the lic ensing phase.  


B/B-UFSAR 1.7-2 REVISION 9 - DECEMBER 2002 TABLE 1.7-1 This Table has been intentionally deleted.  
B/B-UFSAR 1.7-2 REVISION 9 - DECEMBER 2002 TABLE 1.7-1 This Table has been intentionally deleted.  


B/B-UFSAR REVISION 9 - DECEMBER 2002  
B/B-UFSAR REVISION 9 - DECEMBER 2002 Pages 1.7-3 through 1.7-17 have been intentionally deleted.  
 
Pages 1.7-3 through 1.7-17 have been int entionally deleted.
 
B/B-UFSAR  REVISION 9 - DECEMBER 2002
 
Figures 1.1-1 through 1.1-3 have been de leted intentionally.


B/B-UFSAR   REVISION 9 - DECEMBER 2002  
B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.1-1 through 1.1-3 have been deleted intentionally.


Figures 1.2-1 through 1.
B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.2-1 through 1.2-17 have been deleted intentionally.}}
2-17 have been d eleted intentionally.}}

Latest revision as of 15:13, 10 January 2025

Byron/Braidwood Nuclear Stations, Updated Final Safety Analysis Report (Ufsar), Revision 15, Chapter 1.0 Intro General Plant Description
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B/B-UFSAR 1.0-i REVISION 5 - DECEMBER 1994 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT 1.1-1

1.1 INTRODUCTION

1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Site and Environment 1.2-1 1.2.2 Nuclear Steam Supply System 1.2-1 1.2.3 Engineered Safety Features 1.2-2 1.2.4 Emergency Core Cooling System 1.2-3 1.2.5 Control and Instrumentation 1.2-3 1.2.6 Electrical System 1.2-4 1.2.7 Turbine and Auxiliaries 1.2-4 1.2.8 Fuel Handling System 1.2-5 1.2.9 Radioactive Waste Management System 1.2-5 1.2.10 Features of Special Interest 1.2-5 1.2.11 Structures 1.2-6 1.3 COMPARISON TABLES 1.3-1 1.3.1 Comparisons with Similar Facility Designs 1.3-1 1.3.2 Comparison of Final and Preliminary Information 1.3-2 1.3.3 References 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Licensee 1.4-1 1.4.2 Architect-Engineer 1.4-1 1.4.3 Reactor Designer 1.4-1 1.4.4 Constructor 1.4-2 1.4.5 Consultants and Service Organization 1.4-2 1.4.5.1 Security Systems - ETA 1.4-2 1.4.5.2 Dames & Moore 1.4-3 1.4.5.3 HARZA Engineering 1.4-3 1.4.5.4 Murray and Trettel, Inc.

1.4-3 1.4.5.5 Shirmer Engineering Corporation 1.4-3 1.4.5.6 Hyla S. Napadensky 1.4-4 1.4.5.7 NALCO Chemical Company 1.4-4 1.4.5.8 Westinghouse Environmental Systems Department (WESD) 1.4-4 1.4.5.9 Illinois Natural History Survey (INHS) 1.4-5 1.4.5.10 NUS Corporation 1.4-5 1.4.5.11 Eberline Instrument Corporation (EIC) 1.4-5

B/B-UFSAR 1.0-ii REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS (Cont'd)

PAGE 1.4.5.12 Meteorology Research, Inc. (MRI) 1.4-5 1.4.5.13 Illinois State Museum (ISM) 1.4-6 1.4.5.14 Equitable Environmental Health, Inc. (EEH) 1.4-6 1.4.5.15 Espey, Huston & Associates, Inc. (EH & A) 1.4-6 1.4.5.16 University of Wisconsin-Milwaukee (UWM) 1.4-7 1.4.5.17 Aero-Metric Engineering, Inc. (AME) 1.4-7 1.4.5.18 Iowa Institute of Hydraulic Research 1.4-7 1.4.5.19 Babcock and Wilcox International (B&W) 1.4-8 1.4.5.20 Framatome Technologies, Incorporated (FTI) 1.4-8 1.4.5.21 Stone & Webster Engineers and Constructors, Inc, (S&W) 1.4-8 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Programs Required For Plant Operation 1.5-1 1.5.1.1 Core Stability Evaluation 1.5-1 1.5.2 Other Programs Not Required For Plant Operation 1.5-1 1.5.2.1 Fuel Development Program For Operation at High Power Densities 1.5-2 1.5.2.2 Blowdown Forces Program 1.5-2 1.5.2.3 Blowdown Heat Transfer Testing 1.5-2 1.5.3 References 1.5-4 1.6 MATERIAL INCORPORATED BY REFERENCES 1.6-1 1.7 DRAWINGS 1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Drawings for Independent Structural Review 1.7-1

B/B-UFSAR 1.0-iii REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF TABLES NUMBER TITLE PAGE 1.3-1 Plants Using Three-Buttress Containment Design 1.3-3 1.4-1 Exelon Generation Company's Nuclear Power Plants in Service or Under Construction 1.4-9 1.4-2 Nuclear Power Plants Completed or Currently Under Design by Sargent & Lundy 1.4-10 1.4-3 Westinghouse Pressurized Water Reactor Nuclear Power Plants 1.4-11 1.5-1 Blowdown Heat Transfer Phase I Test Parameters 1.5-5 1.5-2 Blowdown Heat Transfer Phase II Test Parameters 1.5-6 1.6-1 Topical Reports Incorporated by Reference 1.6-2 1.7-1 Deleted 1.7-2

B/B-UFSAR 1.0-iv REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 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.

DRAWINGS*

SUBJECT M-1 General Site Plan Units 1 & 2 M-2 Property Development Units 1 & 2 M-5 General Arrangement Roof Plan Units 1 & 2 M-6 General Arrangement Main Floor At El. 451-0 Units 1

& 2 M-7 General Arrangement Mezzanine Floor At El. 426-0 Units 1 & 2 M-8 General Arrangement Grade Floor At El. 401-0 Units 1 & 2 M-9 General Arrangement Floor Plan At El. 383-0 Units 1

& 2 M-10 General Arrangement Basement Floor At El. 364-0 Units 1 & 2 M-11 General Arrangement Floor Plan At El. 346-0 Units 1

& 2 M-12 General Arrangement Radwaste/Service Building Units 1

& 2 M-13 General Arrangement Fuel Handling Building Units 1 &

2 M-14 General Arrangement Section A-A Units 1 & 2 M-15 General Arrangement Section B-B Units 1 & 2 M-16 General Arrangement Section C-C and D-D Units 1 &

2 M-17 General Arrangement Section E-E Units 1 & 2 M-18 General Arrangement Section F-F Units 1 & 2 M-19 General Arrangement Lake Screen House Units 1 & 2 (Braidwood)

M-20 General Arrangement River Screen House Units 1 & 2 M-22 General Arrangement Miscellaneous Plans Units 1 & 2 M-34 P&ID Index and Symbols Units 1 & 2

B/B-UFSAR 1.1-1 REVISION 15 - DECEMBER 2014 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

The Nuclear Regulatory Commission approved the transfer of the facility licenses from Commonwealth Edison (ComEd) Company to Exelon Generation Company, LLC (EGC) on August 3, 2000 (Reference 1). References in the Updated Final Safety Analysis Report (UFSAR) to ComEd, CECo, and Commonwealth Edison have been retained, as appropriate, instead of being changed to EGC to properly preserve the historical context.

This UFSAR is submitted by Exelon Generation Company for nuclear power plants at Byron, Illinois and at Braidwood, Illinois (Drawings M-1 and M-2) in accordance with the requirements of 10 CFR 50.71(e). Each power plant consists of two units having nearly identical nuclear steam supply systems (NSSS) and turbine generators. The main exception is that the original Unit 1 steam generators were replaced by steam generators of a different design. The power plants at the two sites are as nearly identical as site characteristics permit. The bulk of this UFSAR applies to the standardized, non-site-related aspects of the power plants. Sections which describe features specific to the sites are repeated for each site and the applicable station name appears at the top of these pages. Every effort has been made in the preparation of this document to conform to the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", Revision 2, September 1975. The guidance provided in Nuclear Energy Institute (NEI) 98-03, Guidelines for Updating Final Safety Analysis Reports, Revision 1, June 1999, as endorsed by NRC Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10CFR50.71(e), Revision 0, September 1999, is used to comply with the provisions of 10CFR50.71(e).

Each nuclear power plant consists of two nearly identical generating units, and two pressurized water reactor (PWR) (NSSS) and turbine-generator furnished by Westinghouse Electric Corporation (Westinghouse) similar in design concept to several projects recently licensed or currently under review by the NRC (see Section 1.3). Unit 1 contains steam generators supplied by B&W and Unit 2 contains steam generators supplied by Westinghouse.

Westinghouse Electric Corporation, Sargent &

Lundy, and the Commonwealth Edison Company jointly participated in the original design and construction of each unit. The plant is operated by Exelon Generation Company. Sargent & Lundy (S&L) is the architect-engineer for both stations.

Each nuclear steam supply system (NSSS) has been evaluated at a power output of 3672 MWt for the Measurement Uncertainty Recapture (MUR) Power Uprate. The warranted gross and approximate net electrical outputs for the MUR are 1268 MWe and 1241 MWe for Unit 1 and Unit 2, respectively. Safety analyses are evaluated at an NSSS power level of 3672 MWt and a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level of 3648 MWt.

B/B-UFSAR 1.1-1a REVISION 15 - DECEMBER 2014 Specifically, the containment and engineered safety features (ESF) are designed and evaluated for operation at a core thermal power level of 3658 MWt. Accidents (such as loss-of-coolant, steamline break, and other postulated accidents having offsite dose consequences) are also analyzed at a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level 3648 MWt.

B/B-UFSAR 1.1-2 REVISION 9 - DECEMBER 2002 The reactor containments are of post-tensioned concrete construction with a carbon steel liner. Sufficient free volume is provided to contain the energy released in a major accident without need for "pressure suppression" devices. Sargent & Lundy is responsible for containment design.

Byron Station is located in north central Illinois, near the town of Byron and near the Rock River (Drawing M-1). Cooling for the plant is provided by two natural draft cooling towers for non-essential service cooling, and by mechanical draft cooling towers for essential cooling. The fuel loading dates for the two units were November 1984 and November 1986 for Units 1 and 2, respectively. The corresponding dates for commercia1 operation were September 1985 and August 1987.

The Braidwood Station is located in northeastern Illinois, near the town of Braidwood and near the Kankakee River (Drawing M-1).

Cooling for the plant is provided by a large man-made cooling pond of approximately 2500 acres constructed over a previously strip-mined area. Essential service cooling is provided by a 99-acre auxiliary cooling pond which is integral with the main pond. The fuel loading dates for the two units were October 1986 and December 1987 for Units 1 and 2, respectively. The corresponding dates for commercial operation were July 1988 and October 1988.

The standard symbols used on piping and instrument diagrams and other figures in this UFSAR are shown in Drawing M-34.

1.2 REFERENCES

1.

NRC letter, "Braidwood, Byron, Dresden, LaSalle, Quad Cities, and Zion - Orders Approving Transfer of Licenses From Commonwealth Edison Company To Exelon Generation Company, LLC, and Approving Conforming Amendments," dated August 3, 2000

B/B-UFSAR 1.2-1 REVISION 15 - DECEMBER 2014 1.2 GENERAL PLANT DESCRIPTION 1.2.1 Site and Environment The characteristics of the sites and their environs have been investigated to establish bases for determining criteria for storm, flood, and earthquake protection and to evaluate the validity of calculational techniques for the control of routine and accidental releases of radioactive liquids and gases to the environment. Field programs to investigate geology and seismology are completed. Preoperational meteorological programs to provide onsite observations of wind speed and direction have continued since the spring of 1973 at Byron and since the fall of 1973 for Braidwood. Radiological studies of the site environs were initiated at least 18 months prior to commercial operation, with the objective of establishing background radiation levels.

The geography, demography, meteorology, hydrology, geology, and seismology of the two plant sites are discussed in detail in Chapter 2.0.

1.2.2 Nuclear Steam Supply System The nuclear steam supply system (NSSS) consists of a Westinghouse pressurized water reactor and supporting auxiliary systems.

Performance at the calculated steam flow of the NSSS at MUR conditions based on zero percent makeup is as follows:

a.

thermal output of NSSS (MWt) - 3659; b.

thermal output of reactor core (MWt) -3645; c.

steam flow from NSSS (lb/hr) - 16,347,514 for Unit 1/16,280,677 for Unit 2; d.

steam pressure at a steam generator outlet (psia) -

1020.8 for Unit 1 and 902 for Unit 2; e.

maximum moisture content (%) - 0.25%; and f.

feedwater temperature at steam generator inlet (F) -

446.5 for Unit 1 and 447.5 for Unit 2.

The NSSS consists of a reactor and closed reactor coolant loops connected in parallel to the reactor vessel, each loop containing a reactor coolant pump and a steam generator. The NSSS also contains an electrically heated pressurizer and certain auxiliary systems.

B/B-UFSAR 1.2-1a REVISION 7 - DECEMBER 1998 High pressure reactor coolant circulates through the reactor core to remove the heat generated by the nuclear reaction. The heated reactor coolant flows from the reactor vessel to the steam generators (via reactor coolant loop piping). The coolant gives up its heat to the feedwater in the steam generator to generate steam for the turbine generator. The cycle is completed when the reactor coolant is pumped back to the reactor vessel. The entire reactor coolant system is composed of leaktight components to contain the reactor coolant to the system.

B/B-UFSAR 1.2-2 REVISION 11 - DECEMBER 2006 The core is a multiregion type. All fuel assemblies are mechanically identical, although the fuel enrichment is not the same in all assemblies. In a typical initial core loading, three fuel enrichments are used in mechanically identical assemblies.

Fuel assemblies with the highest enrichments are placed in the core periphery, or outer region, and the two groups of lower enrichment fuel assemblies are arranged in a selected pattern in the central region. In subsequent refuelings, one third of the fuel is discharged and fresh fuel is loaded into the outer region of the core. The remaining fuel is arranged in the central two-thirds of the core in such a manner as to achieve optimum power distribution.

Rod cluster control assemblies are used for reactor control and consist of clusters of cylindrical absorber rods. The absorber rods move within guide tubes in certain fuel assemblies. Above the core, each cluster of absorber rods is attached to a spider connector and drive shaft, which is raised and lowered by a drive mechanism mounted on the reactor vessel head. The insertion of the rod cluster control assembly for a reactor trip is by gravity.

The reactor coolant pumps are Westinghouse vertical, single-stage, centrifugal pumps of the shaft-seal type.

The steam generators are B&W vertical U-tube units for Unit 1 and Westinghouse vertical U-tube units for Unit 2. All steam generators contain Inconel tubes. Integral moisture separation equipment reduces the moisture content of the steam.

The reactor coolant piping and all of the pressure-containing surfaces in contact with reactor water are stainless steel. The steam generator tubes and fuel cladding are Inconel and Zircaloy/ZIRLO, respectively. Reactor core internals, including control rod drive shafts, are primarily stainless steel.

An electrically heated pressurizer connected to one reactor coolant loop maintains reactor coolant system pressure during normal operation, limits pressure variations during plant load transients, and keeps system pressure within design limits during abnormal conditions.

Auxiliary system components are provided to charge makeup water into the reactor coolant system, purify reactor coolant, provide chemicals for corrosion inhibition and reactivity control, cool system components, remove decay heat, and provide for emergency safety injection.

1.2.3 Engineered Safety Features The engineered safety features provided for this plant have sufficient redundancy of components and power sources such that

B/B-UFSAR 1.2-3 REVISION 12 - DECEMBER 2008 under the conditions of a loss-of-coolant accident they can maintain the containment integrity and limit personnel exposure to less than 10 CFR 50.67 limits. The engineered safety features incorporated in the design of this plant and the functions they serve are summarized in the following.

a.

The emergency core cooling system injects borated water into the reactor coolant system if coolant is lost. This system limits damage to the core and limits the fission product contamination released into the containment following a postulated loss-of-coolant accident (LOCA).

b.

A steel lined, concrete containment vessel consists of a post-tensioned concrete cylindrical wall and shallow dome, and a conventionally reinforced concrete base. The containment forms a virtually leaktight barrier to prevent the escape of radioactivity.

c.

Reactor containment fan coolers reduce containment temperature and pressure following a postulated loss-of-coolant accident.

d.

A containment spray system is used to reduce containment pressure and to remove iodine and particulate fission products from the containment atmosphere in the event of a loss-of-coolant accident.

e.

The auxiliary feedwater system provides for heat removal from the reactor coolant system by providing makeup water to the steam generator under a variety of postulated conditions.

f.

A combustible gas control system is provided to ensure that the containment atmosphere is mixed following a loss-of-coolant accident. A mixed containment atmosphere prevents local accumulation of combustible or detonable gases that could threaten containment integrity or equipment operating in a local compartment.

1.2.4 Emergency Core Cooling System The emergency core cooling system (ECCS), with passive and active subsystems, is designed to inject borated water into the reactor coolant system (RCS) following a LOCA. This will provide cooling to limit core damage, metal-water reactions, and fission-product release. The ECCS provides long-term postaccident cooling of the core by drawing borated water from the containment sump.

1.2.5 Control and Instrumentation The reactor is controlled by a variety of reactivity coefficients (temperature, pressure, doppler) by control rod cluster motion which is required for load follow transients and for startup and shutdown, and by a soluble neutron absorber, i.e., boron in the

B/B-UFSAR 1.2-4 REVISION 15 - DECEMBER 2014 form of boric acid which is adjusted in concentration to compensate for such effects as fuel consumption and accumulation of fission products.

1.2.6 Electrical System Each unit's main generator is an 1800-rpm, 3-phase, 60-cycle, hydrogen-innercooled unit with water-cooled stator windings and is rated at 1361 MVA at 75 psig gas pressure. Field excitation is provided by a direct shaft-driven brushless excitation system. Two one-half size main step-up transformers deliver power to the 345-kV switchyard.

The station's auxiliary power system consists of system and unit auxiliary transformers; 6900-V, 4160-V, and 480-V switchgear; 480-V motor control centers; 120-Vac instrument buses; and 250-Vdc and 125-Vdc buses.

Two diesel generators are provided for each unit and are available as onsite sources of power (in the event of complete loss of normal a-c power) for operating essential safeguard features. Each diesel generator is capable of supplying required electrical loads for a simultaneous LOCA and loss of offsite power to any one unit.

1.2.7 Turbine and Auxiliaries The turbine for each unit is a four-casing, tandem-compound, six-flow exhaust, 1800-rpm unit with 40-inch last-row blades. There are two combination moisture-separator/steam-reheater assemblies per unit. The turbine-generator for Units 1 have a MUR rating of 1268 MWe gross at 16,347,514 lb/hr steam flow with inlet steam conditions of 1001 psia, 0.36% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. The turbine-generators for Units 2 have a MUR rating of 1241 MWe gross at 16,280,677 lb/hr steam flow with inlet steam conditions of 882 psi, 0.34% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. There are seven stages of feedwater heating for all units.

The turbine is equipped with a redundant fault tolerant Westinghouse Ovation based distributed control system. All control algorithms and processes within the turbine control system are redundant and configured to allow unrestricted turbine operation. This system utilizes a fire-resistant hydraulic fluid to control throttle and governor valve positioning.

B/B-UFSAR 1.2-4a REVISION 11 - DECEMBER 2006 The condenser is of the single-pass deaerating type. There are three parallel strings of feedwater heaters that utilize extraction steam from the low pressure turbines, two parallel strings of feedwater heaters that utilize extraction and exhaust steam from the high pressure turbine, four one-third-sized feedwater condensate and condensate booster pumps, and three one-half-sized feedwater and heater drain pumps. Heater drains from the three highest-pressure feedwater heaters are pumped into the feedwater system; drains from the four lowest-pressure heaters are cascaded to the condenser.

B/B-UFSAR 1.2-5 REVISION 14 - DECEMBER 2012 1.2.8 Fuel Handling System The reactor is refueled with equipment which handles the spent fuel under water from the entire time from leaving the reactor vessel until it is secured in a cask for shipment. Underwater transfer of spent fuel provides a transparent radiation shield and a reliable coolant for decay heat removal.

Fuel handling is performed in the refueling cavity which is flooded for refueling, and the fuel storage pool which is in the fuel building. The two areas are connected by a fuel transfer system which carries the fuel through an opening in the reactor containment.

Spent fuel is removed from the reactor vessel by a refueling machine, placed on the fuel transfer cart conveyor and transferred to the spent fuel storage pool. The fuel is removed from the transfer cart and placed into storage racks. After a suitable decay period, the fuel may be removed from storage and loaded into a shipping cask for removal from the plant.

Refer to Section 9.1.2.3.11 for a description of spent fuel storage and handling using Dry Cask Storage (DCS) system and the Independent Spent Fuel Storage Installation (ISFSI).

All important pumps, piping, and equipment are replicated and capable of being supplied from one of two independent ESF divisions.

1.2.9 Radioactive Waste Management System The radioactive waste system provides equipment necessary to collect, process, and prepare for the disposal of radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation or to transfer the wastes to a vendor-supplied radwaste system.

After collection, depending on chemical composition, liquid wastes may be demineralized and/or filtered. The treated water is discharged at concentrations within the limits of 10 CFR 20.

Sludges and spent demineralizer resins are processed by a vendor-supplied radwaste system for ultimate disposal in an authorized location.

Gaseous wastes are collected from the waste gas header.

Discharge of the gaseous wastes to the environment is controlled to ensure that the offsite dose is as low as reasonably achievable (ALARA).

1.2.10 Features of Special Interest The fundamental concept for the design and construction of the Byron/Braidwood Stations is one of commonality and duplication to the maximum extent permitted by site characteristics. For those features not dictated specifically by site characteristics, identical designs have been employed for the two stations. The concept has been extended to the point where the limiting (i.e.,

B/B-UFSAR 1.2-6 REVISION 9 - DECEMBER 2002 worst case) parameters of the sites are considered in the common design. An example of this is the use of the most restrictive site's seismic building response spectra for the design of systems and components in both plants.

Common plans, drawings, and specifications have been issued for construction at the two sites. Design and construction management for both sites have been conducted by the same major organizations, using the same quality assurance and project management programs. This approach embraces the concept of standardization in nuclear power plant design and construction.

1.2.11 Structures The major structures include a separate and independent containment for each reactor, a common auxiliary building, a common turbine building, a common solid radwaste storage, and administration and service building. General layouts of the plant and interior components' arrangements are shown on Drawings M-5 through M-18 and M-20 and M-22 (Byron), and Drawings M-5 through M-20 and M-22 (Braidwood).

For purposes of design and analysis, structures are designated by Safety Category according to their relation to plant safety. The Safety Category definitions are as follows:

a.

Safety Category I - Those structures important to safety that must be designed to remain functional in the event of the safe shutdown earthquake (SSE) and other design-basis events (including tornado, probable maximum flood, operating basis earthquake (OBE), missile impact, or accident internal to the plant) are designated as Safety Category I.

b.

Safety Category II - Those structures which are not designated as Safety Category I are designated as Safety Category II.

The design criteria and analysis methods for these structures are discussed in Chapter 3.0.

B/B-UFSAR 1.3-1 REVISION 8 - DECEMBER 2000 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs The design is conceptually similar to Exelon Generation Company's Zion Station. Differences in the design of the two plants have been allowed only (1) when dictated by the site characteristics, (2) when the change would result in significant safety improvement, simplification of construction or operation procedures, or cost savings; or (3) as required to comply with appropriate codes and standards, NRC criteria, regulatory guides, and regulations.

The nuclear steam supply system is similar to that of the Zion Station but has a slightly higher power rating. The reactor containments are of the same materials and size as those at the Zion Station, but each has only three buttresses, rather than six as used at Zion. The number of post-tensioning tendons is reduced, and the number of wires per tendon increased, from that used at Zion. The reduced number of buttresses allows for greater separation of penetration areas for redundant safety-related systems. Several plants on which this buttress design has been used are listed in Table 1.3-1.

The polar cranes in the reactor containment are mounted on the containment wall, rather than on the missile barrier as at Zion.

This allows use of a greater area for component laydown in the containment.

Two 100%-capacity containment spray systems are used, rather than the three systems used at Zion. Four containment fan coolers are used, rather than the five used at Zion. The emergency diesel-generator systems for each unit are entirely independent and use two 5500-kW diesel generators per unit. The arrangement of equipment in the common auxiliary building allows greater physical separation of redundant systems and their piping and cables than was possible at Zion.

The Byron Station uses natural draft cooling towers for heat rejection. Zion utilizes once-through cooling. Mechanical draft cooling towers are provided for essential service cooling at Byron.

The Braidwood Station uses a large man-made cooling pond for heat rejection. An auxiliary cooling pond, integral with the main pond, is provided for essential service cooling.

Table 1.3-2 of the FSAR provided the design comparison of the Byron/Braidwood nuclear steam supply system with Comanche Peak, Indian Point 2, South Texas, Sun Desert, W. B. McGuire Nuclear Station, Trojan Nuclear Power Plant, SNUPPS, and the Watts Bar Application. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.

B/B-UFSAR 1.3-2 REVISION 8 - DECEMBER 2000 1.3.2 Comparison of Final and Preliminary Information The Byron/Braidwood Power Plant design was subject to continuing review throughout the construction of the stations. The experience gained at Zion Station and other PWRs was used to enhance equipment reliability and performance. Current design technology was used to upgrade earlier plant design methods.

No significant design changes have been made to the Byron Station or the Braidwood Station which have not been previously reported by amendment to the PSAR, except for the inclusion of 17 x 17 optimized fuel. Table 1.3-3 of the FSAR listed those significant changes reported since the issuance of the Byron and Braidwood Stations Construction Permits. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.

Other changes included the removal of the part length control rods (they are not needed to control Xenon induced axial instabilities), the enlargement of spent fuel capacity, the use of more corrosion-resistant materials in the steam generators and moisture steam separators, improved equipment packaging to do a reactor refueling in a shorter time period, an upgraded design for the reactor coolant pump seals, and replacement steam generators for Unit 1. These concepts are described in later chapters.

1.3.3 References

1. Exelon Generation Company, "Byron/Braidwood Stations Fire Protection Report in Response to Appendix A of BTP APCSB 9.5-1,"

(current amendment).

B/B-UFSAR 1.3-3 TABLE 1.3-1 PLANTS USING THREE-BUTTRESS CONTAINMENT DESIGN PLANT/UTILITY DATE OF OPERATION Arkansas Nuclear One Arkansas Power & Light Co.

5-21-74 Millstone-2 Northeast Utilities 8-1-75 Rancho Seco Sacramento Municipal Utility District 8-16-74 Trojan Portland General Electric Co.

11-21-75 J.M. Farley-1 Alabama Power Co.

6-25-77

B/B-UFSAR 1.4-1 REVISION 8 - DECEMBER 2000 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 Licensee Exelon Generation Company is the Licensee for the Byron Station, Units 1 and 2, which is located in Rockvale Township, Ogle County, approximately 4 miles south of Byron, Illinois, and for Units 1 and 2 of the Braidwood Station, which is located in Reed Township, Will County, approximately 6 miles southwest of Wilmington, Illinois. The Licensee is responsible for the design, construction, and operation of the nuclear power plants.

Commonwealth Edison supplies electrical service to an area of 13,000 square miles with a population of approximately 8 million persons, located primarily in the northern third of Illinois.

Dresden 1, Commonwealth Edison's first nuclear generating station, went into commercial service during August 1960, and has produced more than 10 billion kWh. Additional nuclear units in service or under construction are listed in Table 1.4-1.

1.4.2 Architect-Enqineer For the work covered by this application, Sargent & Lundy (S&L) has been retained as the design consultants. The Licensee has employed Sargent & Lundy for power plant design work for over 80 years.

Sargent & Lundy is an independent consulting engineering organization founded in Chicago, in 1891. For over three-quarters of a century, the firm has specialized exclusively in the design of generation, transmission, distribution, and utilization of steam and electric power and related facilities.

The firm has provided the complete engineering services for more than 600 turbine-generator units with a total capacity of 53,000,000 kW. Of this total, some 9,800,000 kW is in nuclear generating capacity. Table 1.4-2 lists the nuclear plants completed by or currently under design by Sargent & Lundy.

1.4.3 Reactor Designer Westinghouse has designed, developed, and manufactured nuclear power facilities since the 1950s, beginning with the world's first large central station nuclear power plant (Shippingport),

which started producing power in 1957. Completed or contracted

B/B-UFSAR 1.4-2 REVISION 5 - DECEMBER 1994 commercial nuclear capacity totals were in excess of 98,000 MWe. Westinghouse pioneered new nuclear design concepts, such as chemical shim control of reactivity and the rod cluster control concept, throughout the last two decades. Westinghouse manufacturing facilities include the largest commercial nuclear fuel fabrication facility in the world and the world's most modern heat transfer equipment production facility, as well as other facilities producing nuclear steam supply system (NSSS) components. Table 1.4-3 lists all Westinghouse pressurized water reactor (PWR) plants to date, including those plants under construction or on order at the time of the Byron/Braidwood application.

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute have contracted with Westinghouse for research into NSSS-related activities. Westinghouse experience was also utilized by the NRC and Metropolitan Edison immediately following the Three Mile Island Unit 2 accident and the corporation continues to participate with the Westinghouse Owner's Group of utilities in addressing the NRC action plan and other operations improvements.

1.4.4 Constructor Construction coordination of all activities at the site was under the supervision of the Commonwealth Edison's Station Construction Department. The department exercises site managerial functions as discussed in Chapter 17.0 of the UFSAR.

The Station Construction Department was the constructor for Zion Station. This department has coordinated the construction activities for almost all of Commonwealth Edison's existing power plants. It was also the construction coordinator for La Salle County Station.

1.4.5 Consultants and Service Organization 1.4.5.1 Security System - ETA The design of the physical security system and the administrative controls was performed by ETA, Inc.

ETA personnel have had varied and in-depth experience in the design, safety analysis, and environmental review of nuclear power plants and related facilities as well as in the management and organization of security systems. They are very familiar with the details of the current generation of light water reactors and, in particular, those critical areas and components of the plants which might be the most vulnerable to sabotage.

They are also familiar with the current regulations and guidelines of the NRC that define the required performance and objectives of a security system for licensed activities.

B/B-UFSAR 1.4-3 1.4.5.2 Dames & Moore The independent consulting firm of Dames & Moore was employed to conduct studies relating to the geology, seismology, and groundwater hydrology at both sites. The firm also conducted preconstruction baseline studies, including wildlife surveys as well as soil and vegetation analyses.

Having performed environmental studies for approximately 30 nuclear power plant sites, Dames & Moore is a recognized authority in the field of environmental engineering of nuclear power plants.

1.4.5.3 HARZA Engineering HARZA was employed in the design of the water treatment facilities at both stations.

HARZA has been involved with a variety of technical studies for at least ten nuclear power stations. Among these studies have been facility design, review of design and structure, hydrology, and groundwater. In addition, HARZA Engineering has designed some of the largest hydroelectric projects in the world, including major concrete structures and earthfilled dams.

1.4.5.4 Murray and Trettel, Inc.

Murray and Trettel (M&T) is an environmental consulting firm which, since 1960, has provided significant meteorological input to both preoperational and operational phases of meteorological programs for nuclear power stations. M&T has also provided meteorological input to a wide variety of air pollution and environmental problems as well as allied control technique programs.

Murray & Trettel provided meteorological data for both stations by implementation of an onsite measurement program incorporating a tower for elevation measurements.

1.4.5.5 Shirmer Engineering Corporation Shirmer Engineering is a firm of consulting fire protection engineers. The firm has done work on 17 Department of Energy nuclear fuel production and laboratory facilities, as well as for numerous nuclear power stations for Sargent & Lundy.

Shirmer Engineering has also performed services for many fossil units.

Shirmer Engineering provided evaluation of the fire protection systems at both stations and assisted in the preparation of the Byron/Braidwood Fire Protection Report.

B/B-UFSAR 1.4-4 1.4.5.6 Hyla S. Napadensky Ms. Napadensky was retained to help evaluate the probability of an accidental explosion occurring on a train carrying explosives in the vicinity of the Braidwood Station.

Ms. Napadensky is the Manager of Fire Safety Research at the IIT Research Institute of the Illinois Institute of Technology.

Ms. Napadensky has directed analytical and experimental research in the areas of explosion effects, hazards and risk analysis, safety of chemical systems, explosives and propellant sensitivity, and initiation mechanisms during her 17 years with IIT Research Institute.

1.4.5.7 NALCO Chemical Company The NALCO Chemical Company (formerly Industrial Bio-Test, Inc.)

consisted of two divisions, Industrial Bio-Test Laboratories, and NALCO Environmental Sciences, which conduct studies relating to toxicology and ecological sciences, respectively. The Environmental Science Division includes seven subdivisions: (1) aquatic biology, (2) fisheries and field operations, (3) water and wastewater chemistry, (4) radiochemistry, (5) air sciences and data processing, (6) land and plant sciences, and (7) environmental physiology.

As a technical consultant on the Braidwood project, the NALCO Chemical Company provided a clam bed mapping survey in the area of the station's intake and discharge structures located on the Kankakee River.

1.4.5.8 Westinghouse Environmental Systems Department (WESD)

WESD, established as a department of the Westinghouse Power Systems Company in 1969, consisted of environmental scientists and engineers experienced in the areas of aquatic and terrestrial biology and ecology; geology; limnology; environmental chemistry and physics; physical oceanography, meteorology and climatology, radiology, public health aspects of pollutant emissions, and systems engineering and integration.

WESD conducts broad environmental surveys, environmental program planning and data interpretation, and provides recommended action programs for meeting federal, state, and local environmental quality regulations. As a technical consultant on the Braidwood project, WESD staff biologists conducted a 2-year baseline study of the Braidwood Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Kankakee River in the vicinity of the site were determined, and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation in the biotic communities of the site were predicted.

B/B-UFSAR 1.4-5 1.4.5.9 Illinois Natural History Survey (INHS)

The Illinois Natural History Survey (INHS), which has its beginnings almost 120 years ago, is a division of the State Department of Registration and Education and provides services to farmers, homeowners, sportsmen, and all other citizens of Illinois as well as to industries. INHS cooperates in biological research with the Illinois Department of Agriculture, Conservation, and Public Health; the University of Illinois, Southern Illinois University, and other educational institutions; various research branches of the federal government; and other agricultural, conservation, municipal, and business organizations throughout the state.

INHS aquatic biologists were involved in a 4-year study of the Kankakee River and Horse Creek near Custer Park, Illinois. The purpose of the study is to obtain biological, physical, and chemical data which will be used to evaluate any effects of the construction and operation of the Braidwood Station and its associated cooling lake on the biota and water quality of the Kankakee River and Horse Creek. The station's cooling pond will use the Kankakee River as a source of water for both intake and discharge purposes.

1.4.5.10 NUS Corporation NUS Corporation is a consulting engineering, research, and testing firm specializing in environmental and energy systems engineering, systems analysis, design engineering, management consulting, and training programs related to these areas. NUS has provided advice and professional guidance to utility, industrial, and government clients throughout the United States and in a number of foreign countries.

As a technical consultant on the Braidwood project, NUS was involved in a study to determine the adequacy of the station's ultimate heat sink.

1.4.5.11 Eberline Instrument Corporation (EIC)

Eberline Instrument Corporation (EIC) has provided radiation measurement equipment, comprehensive radiation protection services, and analytical laboratory services to the nuclear industry since 1953.

As a technical consultant on the Byron/Braidwood projects EIC performed preoperational environmental radiological baseline studies on and around the site.

1.4.5.12 Meteorology Research, Inc. (MRI)

Meteorology Research, Inc. (MRI) is an environmental consulting firm which, since 1951, has provided meteorological and air

B/B-UFSAR 1.4-6 REVISION 1 - DECEMBER 1989 quality instruments and services to all aspects of industry in the solution of weather-related problems. These range from environmental impact assessments of existing or proposed airports and other major developments to problems of warehousing and marketing seasonal consumer goods. Of particular interest is the influence the local topography has on temperatures and winds.

MRI provided meteorological data from 1973 through mid-1975 for Byron and Braidwood Stations by implementation of an onsite meteorological measurement program.

1.4.5.13 Illinois State Museum (ISM)

The Illinois State Museum conducts archaeological investigations throughout the state of Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing and excavating, and laboratory analyses of datifacts, pollen, and soils.

As a technical consultant on the Braidwood project, ISM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Braidwood Station and pipeline corridor property.

1.4.5.14 Equitable Environmental Health, Inc. (EEH)

Equitable Environmental Health, Inc. (EEH), successor to Environmental Analysts, Inc./Tabershaw-Cooper Associated, Inc.,

is a multidisciplinary firm that offers the consulting services of medical professionals, scientists, engineers, economists, and technical support personnel in all areas of environmental health and economics.

EEH staff biologists conducted a 2-year baseline study of the Byron Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Rock River in the vicinity of the site were determined and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation on the biotic communities of the site were predicted, and data were provided for a benefit-cost analysis of the project.

1.4.5.15 Espey, Huston & Associates, Inc. (EH & A)

Espey, Huston & Associates, Inc. (EH & A) is a consulting firm addressing the environmental problems associated with industrial and urban development. EH & A professionals cover a broad range of expertise including civil engineering, environmental engineering, mathematics and computer science, and all phases of aquatic, estuarine, and terrestrial ecology.

B/B-UFSAR 1.4-7 REVISION 1 - DECEMBER 1989 As a technical consultant on the Byron project, EH & A conducted the construction phase terrestrial and aquatic monitoring programs.

1.4.5.16 University of Wisconsin-Milwaukee (UWM)

The University of Wisconsin-Milwaukee under Dr. Elizabeth Benchley of the Dept. of Anthropology, conducts archaeological investigations throughout Wisconsin and northern Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing, and lab analysis of datifacts, pollen, and soils.

As a technical consultant on the Byron project, UWM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Byron Station and pipeline corridor property. Also, UWM conducted archaeological investigations on the Byron transmission line right-of-ways.

1.4.5.17 Aero-Metric Engineering, Inc. (AME)

Aero-Metric Engineering, Inc., founded in 1969, is based in Sheboygan, Wisconsin. The staff was made up of over 50 technical photogrammetric personnel, many having professional engineer and/or survey registration. AME's capabilities allow for a complete range of precision photogrammetric services, including aerial photography, mapping, and multiple survey skills.

As a technical consultant on the Byron project, AME will be providing annual aerial infra-red photographs.

1.4.5.18 Iowa Institute of Hydraulic Research The Iowa Institute of Hydraulic Research, formally organized in 1931, is a Division of the University of Iowa's College of Engineering. The Institute staff exceeded 80 in number and was comprised of a professional staff with Ph.Ds in the areas of Civil Engineering, Mechanical Engineering, Physics, Mechanics and Hydraulics, and Aeronautical Engineering, with most of these personnel holding joint academic appointments in the College of Engineering's Division of Energy Engineering. The Institute of Hydraulic Research conducts programs of fundamental research and advanced design and analysis in the areas of environmental pollution, bioengineering, naval hydrodynamics, river mechanics, ice hydraulics, hydrology, water resources, hydraulic structures, fluid mechanics, advanced instrumentation and data-handling techniques for fluids research, and mathematical modeling of watersheds and hydrology.

B/B-UFSAR 1.4-8 REVISION 9 - DECEMBER 2002 As a technical consultant on the Braidwood project, the Institute conducted a thermal evaluation to determine the adequacy of the ultimate heat sink.

1.4.5.19 Babcock and Wilcox International (B&W)

B&W is located in Cambridge, Ontario, Canada. B&W has fabricated fossil-fueled boiler components for over 100 years and has fabricated nuclear system components since the late 1950's. B&W has supplied replacement steam generators for Byron Unit 1 and Braidwood Unit 1.

1.4.5.20 Framatome Technologies, Incorporated (FTI)

FTI is located in Lynchburg, Virginia and has been providing services to the electric power industry for over four decades.

FTI engineering services include the necessary expertise, experience, and NRC-approved computer codes and methodologies to support the transient analysis of the Unit steam generators.

1.4.5.21 Stone & Webster Engineers and Constructors, Inc. (S&W)

S&W is located in Boston, Massachusetts and has been providing services to the electric power industry for over 100 years. S&W has provided balance-of-plant design-engineering support services in support of the power uprate of the Byron and Braidwood units.

B/B-UFSAR 1.4-9 REVISION 8 - DECEMBER 2000 TABLE 1.4-1 EXELON GENERATION COMPANY'S NUCLEAR POWER PLANTS IN SERVICE OR UNDER CONSTRUCTION UNIT NOMINAL GROSS1 RATING (MWe)

SCHEDULED COMMERCIAL SERVICE DATE Dresden 1 210 1960 Dresden 2 850 1972 Dresden 3 850 1972 Quad-Cities 1 850 1972 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 La Salle 1 1122 1978 La Salle 2 1122 1979 Byron 1 1175 1985 Byron 2 1175 1987 Braidwood 1 1175 1988 Braidwood 2 1175 1988 1Note that this is a gross rating, not a net rating.

B/B-UFSAR 1.4-10 REVISION 1 - DECEMBER 1989 TABLE 1.4-2 NUCLEAR POWER PLANTS COMPLETED OR CURRENTLY UNDER DESIGN BY SARGENT & LUNDY UNIT NOMINAL GROSS2 RATING (MWe)

YEAR OF POWER OPERATION EBWR 5

1956 Elk River 22 1962 La Crosse 60 1967 SEFOR 20 (MWt) 1969 Dresden 2 850 1969 Dresden 3 850 1971 Quad-Cities 1 850 1971 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 Fort St. Vrain, Unit 1 330 1973 Enrico Fermi, Unit 2 1200 1988 La Salle County Station, Unit 1 1122 1979 La Salle County Station, Unit 2 1122 1980 Byron Station, Unit 1 1175 1985 Byron Station, Unit 2 1175 1987 Braidwood Station, Unit 1 1175 1988 Braidwood Station, Unit 2 1175 1988 Clinton Power Station, Unit 1 992 1981 Kaiseraugst 992 1982 2Note that this is a gross rating, not a net rating.

B/B-UFSAR 1.4-11 TABLE 1.4-3 WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTS PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Shippingport Duquesne Light Company; Energy Research & Development Administration Pennsylvania 1957 90 4

Yankee-Rowe Yankee Atomic Electric Company Massachusetts 1961 175 4

Trio Vercellese (Enrico Fermi)

Ente Nazionale per L'Energia Elettrica (ENEL)

Italy 1965 260 4

Chooz (Ardennes)

Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA)

France 1967 305 4

San Onofre Unit 1 Southern California Edison Co.;

San Diego Gas and Electric Co.

California 1968 450 3

Haddam Neck (Connecticut Yankee)

Connecticut Yankee Atomic Power Company Connecticut 1968 575 4

Jose Cabrera-Zorita Union Electrica, S.A.

Spain 1969 153 1

Beznau Unit 1 Nordostschweizerische Krafwerke AG (NOK)

Switzerland 1969 350 2

Robert Emmett Ginna Rochester Gas and Electric Corporation New York 1970 490 2

Mihama Unit 1 The Kansai Electric Power Company, Inc.

Japan 1970 320 2

Point Beach Unit 1 Wisconsin Electric Power Co.;

Wisconsin Michigan Power Co.

Wisconsin 1970 497 2

H. B. Robinson Unit 2 Carolina Power and Light Co.

South Carolina 1971 707 3

B/B-UFSAR 1.4-12 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beznau Unit 2 Nordostschweizerische Kraftwerke AG (NOK)

Switzerland 1972 350 2

Point Beach Unit 2 Wisconsin Electric Power Co.;

Wisconsin Michigan Power Co.

Wisconsin 1972 497 2

Surry Unit 1 Virginia Electric and Power Co.

Virginia 1972 822 3

Turkey Point Unit 3 Florida Power and Light Co.

Florida 1972 745 3

Indian Point Unit 2 Consolidated Edison Company of New York, Inc.

New York 1973 873 4

Prairie Island Unit 1 Northern States Power Company Minnesota 1973 530 2

Turkey Point Unit 4 Florida Power and Light Co.

Florida 1973 745 3

Surry Unit 2 Virginia Electric and Power Co.

Virginia 1973 822 3

Zion Unit 1 Exelon Generation Company Illinois 1973 1050 4

Kewaunee Wisconsin Public Service Corp.;

Wisconsin Power and Light Co.;

Madison Gas and Electric Co.

Wisconsin 1974 560 2

Prairie Island Unit 2 Northern States Power Company Minnesota 1974 530 2

Takahama Unit 1 The Kansai Electric Power Company, Inc.

Japan 1974 781 3

Zion Unit 2 Exelon Generation Company Illinois 1974 1050 4

B/B-UFSAR 1.4-13 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Doel Unit 1 Indivision Doel Belgium 1975 390 2

Doel Unit 2 Indivision Doel Belgium 1975 390 2

Donald C. Cook Unit 1 Indiana and Michigan Electric Company (AEP)

Michigan 1975 1060 4

Ringhals Unit 2 Statens Vattenfallsverk (SSPB)

Sweden 1975 822 3

Almaraz Unit 1 Unit Electrica, S.A.;

Compania Sevillana de Electricidad, S.A.;

Hidroelectrica Espanola, S.A.

Spain 1976 902 3

Beaver Valley Unit 1 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Company Pennsylvania 1976 852 3

Diablo Canyon Unit 1 Pacific Gas and Electric Co.

California 1976 1084 4

Indian Point Unit 3 Consolidated Edison Company of New York, Inc.

New York 1976 965 4

Lemoniz Unit 1 Iberduero, S.A.

Spain 1976 902 3

Salem Unit 1 Public Service Electric and Gas Company; Exelon Generation Company; Atlantic City Electric Co.;

Delmarva Power and Light Co.

New Jersey 1976 1090 4

B/B-UFSAR 1.4-14 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Trojan Portland General Electric Co.;

Eugene Water and Electric Board; Pacific Power and Light Company Oregon 1976 1130 4

Almaraz Unit 2 Union Electrica, S.A.;

Compania Sevillana de Electricidad, S.A.;

Hidroelectrica Espanola, S.A.

Spain 1977 902 3

Asco Unit 1 Fuerzas Electricas de Cataluna, S.A. (FESCA)

Spain 1977 902 3

Diablo Canyon Unit 2 Pacific Gas and Electric Co.

California 1977 1106 4

Joseph M. Farley Unit 1 Alabama Power Company Alabama 1977 829 3

Ko-Ri Unit 1 Korea Electric Power Co., Ltd.

Korea 1977 564 2

North Anna Unit 1 Virginia Electric and Power Co.

Virginia 1977 898 3

North Anna Unit 2 Virginia Electric and Power Co.

Virginia 1977 898 3

Ohi Unit 1 The Kansai Electric Power Co., Inc.

Japan 1977 1122 4

Ohi Unit 2 The Kansai Electric Power Co., Inc.

Japan 1977 1122 4

Ringhals Unit 3 Statens Vattenfallsvert (SSPB)

Sweden 1977 900 3

Sequoyah Unit 1 Tennessee Valley Authority Tennessee 1977 1148 4

Angra dos Reis Unit 1 Furnas-Centrais Electricas, S.A.

Brazil 1978 626 2

B/B-UFSAR 1.4-15 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Asco Unit 2 Fuerzas Electricas de Cataluna, S.A. (FESCA);

Empresa Nacional Hidroelectrica del Ribagorzana, S.A. (ENHER);

Fuerzas Hidroelectricas del Segre, S.A.;

Hidroelectrica de Cataluna, S.A.

Spain 1978 902 3

Donald C. Cook Unit 2 Indiana and Michigan Electric Company (AEP)

Michigan 1978 1060 4

Lemoniz Unit 2 Iberduero, S.A.

Spain 1978 902 3

Sequoyah Unit 2 Tennessee Valley Authority Tennessee 1978 1148 4

Watts Bar Unit 1 Tennessee Valley Authority Tennessee 1978 1177 4

William B. McGuire Unit 1 Duke Power Company North Carolina 1978 1180 4

Joseph M. Farley Unit 2 Alabama Power Company Alabama 1979 829 3

Krsko Savske Elektrarne, Ljubljana, Slovenia, Elektroprivreda, Zagreb, Croatia Yugoslavia 1979 615 2

Ringhals Unit 4 Statens Vattenfallsvert (SSPD)

Sweden 1979 900 3

B/B-UFSAR 1.4-16 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Salem Unit 2 Public Service Electric and Gas Company; Exelon Generation Company Atlantic City Electric Co.;

Delmarva Power and Light Co.

New Jersey 1979 1115 4

Virgil C. Summer South Carolina Electric and Gas Company South Carolina 1979 900 3

Watts Bar Unit 2 Tennessee Valley Authority Tennessee 1979 1177 4

William B. McGuire Unit 2 Duke Power Company North Carolina 1979 1180 4

Byron Unit 1 Exelon Generation Company Illinois 1981 1120 4

Comanche Peak Unit 1 Texas Utilities Generating Co.

Texas 1980 1150 4

Seabrook Unit 1 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1980 1200 4

South Texas Project Unit 1 Houston Lighting and Power Co.;

Central Power and Light Co.;

City Public Service of San Antonio; City of Austin, Texas Texas 1980 1250 4

B/B-UFSAR 1.4-17 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beaver Valley Unit 2 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Co.;

Cleveland Electric Illuminating Company; Toledo Edison Company Pennsylvania 1981 852 3

Braidwood Unit 1 Exelon Generation Company Illinois 1981 1120 4

Callaway Unit 1 SNUPPS - Union Electric Co.

Missouri 1981 1150 4

Catawba Unit 1 Duke Power Company South Carolina 1981 1153 4

Jamesport Unit 1 Long Island Lighting Company New York 1981 1150 4

Ko-Ri Unit 2 Korea Electric Power Co., Ltd.

Korea 1981 605 2

NORCO Puerto Rico Water Resources Authority Puerto Rico 583 2

Braidwood Unit 2 Exelon Generation Company Illinois 1982 1120 4

Byron Unit 2 Exelon Generation Company Illinois 1982 1120 4

Catawba Unit 2 Duke Power Company South Carolina 1982 1153 4

Comanche Peak Unit 2 Texas Utilities Generating Co.

Texas 1982 1150 4

Marble Hill Unit 1 Public Service Company of Indiana, Inc.;

Northern Indiana Public Service Company Indiana 1982 1150 4

B/B-UFSAR 1.4-18 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Millstone Unit 3 Northeast Nuclear Energy Co.

Connecticut 1982 1156 4

Seabrook Unit 3 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1982 1200 4

South Texas Project Unit 2 Houston Lighting and Power Co.;

Central Power and Light Co.;

City Public Service of San Antonio; City of Austin, Texas Texas 1982 1250 4

Taiwan Unit 5 Taiwan Power Company Taiwan 1982 950 3

Wolf Creek Unit 1 SNUPPS - Kansas Gas and Electric Company; Kansas City Power and Light Company Kansas 1982 1150 4

Alvin W. Vogtle Unit 1 Georgia Power Company Georgia 1983 1113 4

Callaway Unit 2 SNUPPS - Union Electric Company Missouri 1983 1150 4

NEP-1 New England Power Company 1983 1150 4

Fort Calhoun Unit 2 Omaha Public Power District; Nebraska Public Power District Nebraska 1983 1150 4

Jamesport Unit 2 Long Island Lighting Company New York 1983 1150 4

Sears Island Central Maine Power Company Maine 1200 4

B/B-UFSAR 1.4-19 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Taiwan Unit 6 Taiwan Power Company Taiwan 1983 950 3

Alvin W. Vogtle Unit 2 Georgia Power Company Georgia 1984 1113 4

Marble Hill Unit 2 Public Service Company of Indiana, Inc.;

Northern Indiana Public Service Company Indiana 1984 1150 4

Shearon Harris Unit 1 Carolina Power and Light Co.

North Carolina 1984 900 3

Sterling SNUPPS - Rochester Gas and Electric Corporation; Central Hudson Gas and Electric Corporation; Niagara Mohawk Power Corporation; Orange and Rockland Utilities, Inc.

New York 1984 1150 4

Atlantic Unit 1 (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Co.;

Jersey Central Power and Light Company New Jersey 1985 1150 4

NEP-2 New England Power Company 1985 1150 4

South Dade Unit 1 Florida Power and Light Co.

Florida 1985 1150 4

Sundesert Unit 1 San Diego Gas and Electric Co.

California 1985 950 3

B/B-UFSAR 1.4-20 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Tyrone Unit 1 SNUPPS - Northern States Power Company Wisconsin 1985 1150 4

Shearon Harris Unit 2 Carolina Power and Light Co.

North Carolina 1986 900 3

South Dade Unit 2 Florida Power and Light Co.

Florida 1986 1150 4

Atlantic No. (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Co.;

Jersey Central Power and Light Company New Jersey 1987 1150 4

Shearon Harris Unit 4 Carolina Power and Light Co.

North Carolina 1988 900 3

Sundesert Unit 2 San Diego Gas and Electric Co.

California 1988 950 3

Sayago Unit 1 Iberduero, S.A.

Spain 1980's 1000 3

Sayago Unit 4 Iberduero, S.A.

Spain 1980's 1000 3

Shearon Harris Unit 3 Carolina Power and Light Co.

North Carolina 1990 900 3

Unassigned Unit 1 (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1990 1150 4

Unassigned Unit 2 (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1992 1150 4

B/B-UFSAR 1.5-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION The design of the Byron/Braidwood units is based upon proven concepts which have been developed and successfully applied to the design of pressurized water reactor systems. There are currently no areas of research and development which are required for operation of this plant.

At the time of issuance of construction permits for the Byron/

Braidwood units, the Preliminary Safety Analysis Report (PSAR) and the standard design report which it referenced, RESAR-3, identified certain research and development programs which were incomplete. These programs, which have been successfully completed, have provided technical information which has been used either to demonstrate the safety of design, more sharply define margins of conservatism, or lead to design improvements.

Reference 1 presents descriptions of those safety-related research and development programs which have been carried out for, by, or in conjunction with Westinghouse Nuclear Energy Systems, and which are applicable to Westinghouse pressurized water reactors. The discussion which follows documents the completion of the construction permit stage research programs.

1.5.1 Programs Required for Plant Operation Two programs were identified as required for plant design and operation in the PSAR:

a. core stability evaluation and
b. fuel rod burst program.

Both programs are complete. The fuel rod burst program was completed at the time of the PSAR. The core stability evaluation program was not. A discussion of the core stability evaluation program follows.

1.5.1.1 Core Stability Evaluation The program to establish means for the detection and control of potential xenon oscillations and for the shaping of the axial power distribution for improved core performance has been satisfactorily completed. See item 1, Reference 2, for a further discussion of the tests and results.

1.5.2 Other Programs Not Required for Plant Operation The following programs were not complete at the time of the PSAR but are now satisfactorily complete.

B/B-UFSAR 1.5-2 1.5.2.1 Fuel Development Program for Operation at High Power Densities The program to demonstrate the satisfactory operation of fuel at high burnup and power densities has been satisfactorily completed. See item 8, Reference 2, for a further discussion of the program and its results.

1.5.2.2 Blowdown Forces Program Westinghouse has completed BLODWN-2, an improved digital computer program for the calculation of local fluid pressures, flows and density transients in the primary coolant systems during a LOCA.

BLODWN-2 is used to evaluate the effects of blowdown forces in this application. Refer to item 15 in Reference 4 for a further discussion of the tests and results.

1.5.2.3 Blowdown Heat Transfer Testing (Formerly Titled Delayed Departure From Nucleate Boiling)

The NRC Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Power Reactors was issued in Section 50.46 of 10 CFR 50 on December 28, 1973. It defines the basis and conservative assumptions to be used in the evaluation of the performance of emergency core cooling systems (ECCS).

Westinghouse believes that some of the conservatism of the criteria is associated with the manner in which transient DNB phenomena are treated in the evaluation models. Transient critical heat flux data presented at the 1972 specialists meeting of the Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be delayed under transient conditions.

To demonstrate the conservatism of the ECCS evaluation models, Westinghouse initiated a program to experimentally simulate the blowdown phase of a LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blowdown Heat Transfer Program, which was started early in 1976. Testing was completed in 1979. A DNB correlation developed by Westinghouse from these test results is used in the ECCS analyses for Byron/Braidwood.

Objective The objective of the blowdown heat transfer test was to determine the time that DNB occurs under LOCA conditions. This information was used to confirm a new Westinghouse transient DNB correlation.

The steady-state DNB data obtained from 15x15 and 17x17 test programs was used to assure that the geometrical differences between the two fuel arrays is correctly treated in the transient correlations.

B/B-UFSAR 1.5-3 Program The program was divided into two phases. The Phase I tests started from steady-state conditions, with sufficient power to maintain nucleate boiling throughout the bundle, and progressed through controlled ramps of decreasing test section pressure or flow initiated DNB. By applying a series of controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pressures relevant to a PWR blowdown.

Phase I provided separate-effects data for heat transfer correlation development.

Typical parameters used for Phase I testing are shown in Table 1.5-1.

Phase II simulated PWR behavior during a LOCA to permit definition of the time delay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after establishment of typical steady-state operating conditions. The fluid transient was then initiated, and the rod power decay was programmed in such a manner as to simulate the actual heat input of fuel rods. The test was terminated when the heater rod temperatures reached a predetermined limit.

Typical parameters used for Phase II testing are shown in Table 1.5-2.

Test Description The experimental program was conducted in the J-Loop at the Westinghouse Forest Hills Facility with a full length 5x5 rod bundle simulating a section of a 15x15 fuel assembly to determine DNB occurrence under LOCA conditions.

The heater rod bundles used in this program were internally-heated rods, capable of a maximum linear power of 18.8 kW/ft, with a total power of 135 kW (for extended periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribution. Each rod was adequately instrumented with a total of 12 clad thermocouples.

Results The experiments in the DNB facility resulted in cladding temperature and fluid properties measured as a function of time throughout the blowdown range from 0 to 20 seconds.

Facility modifications and installation of the initial test bundle were completed. A series of shakedown tests in the

B/B-UFSAR 1.5-4 REVISION 1 - DECEMBER 1989 J-Loop were performed. These tests provided data for instrumentation calibration and check-out, and provided information regarding facility control and performance. Initial program tests were performed during the first half of 1975.

Under the sponsorship of EPRI, testing was reinitiated during 1976 on the same test bundle. The testing was terminated in November 1976 and plans were made for a new test bundle and further testing during 1978-1979. These tests were completed in December of 1979.

1.5.3 References

1.

F. T. Eggleston, "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, October 1978.

2.

F. T. Eggleston, "Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8768.

Spring 1976 Edition.

3.

"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8458. Fall 1977 Edition.

4.

"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8004. Fall 1972 Edition.

B/B-UFSAR 1.5-5 TABLE 1.5-1 BLOWDOWN HEAT TRANSFER PHASE I TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 1250 to 2250 psia Test section mass velocity 1.12 to 2.5x106 lb/hr-ft2 Core inlet temperature 550° F to 600° F Maximum heat flux 306,000 to 531,000 Btu/hr-ft2 TRANSIENT RAMP CONDITIONS Pressure decrease 0 to 350 psia/sec and subcooled depressurization from 2250 psia Flow decrease 0 to 100%/sec Inlet enthalpy constant

B/B-UFSAR 1.5-6 TABLE 1.5-2 BLOWDOWN HEAT TRANSFER PHASE II TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 2250 psia Test section mass velocity 2.5x106 lb/hr-ft2 Inlet coolant temperature 545° F Maximum heat flux 531,000 Btu/hr-ft2 TRANSIENT CONDITIONS Simulated break Double-ended cold leg guillotine breaks

B/B-UFSAR 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCES Table 1.6-1 lists topical reports which provide information additional to that provided in this UFSAR and which have been filed separately with the Nuclear Regulatory Commission (NRC) in support of this and similar applications.

A legend to the review status code letters follows:

A

- NRC review complete; NRC acceptance letter issued.

AE

- NRC accepted as part of the Westinghouse Emergency Core Cooling System (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.

B

- Submitted to the NRC as background information; not undergoing formal NRC review.

O

- On file with NRC; older generation report with current validity; not actively under formal NRC review.

U

- Actively under formal NRC review.

B/B-UFSAR 1.6-2 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE REPORT REFERENCE SECTION(S)

REVIEW STATUS "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," WCAP-2048, July 1962 4.3 0

"Single Phase Local Boiling and Bulk Boiling Pressure Drop Correlations," WCAP-2850 (Proprietary), April 1966 and WCAP-7916 (Non-Proprietary), June 1972 4.4 0

"In-Pile Measurement of UO2 Thermal Conductivity," WCAP-2923, 1966 4.4 0

"Hydraulic Tests of the San Onofre Reactor Model," WCAP-3269-8, June 1964 4.4 0

"LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM - 7094,"

WCAP-3269-26, September 1963 4.3, 4.4 15.0, 15.4 0

"Saxton Core II Fuel Performance Evaluation,"

WCAP-3385-56, Part II, "Evaluation of Mass Spectrometric and Radiochemical Analyses of Irradiated Saxton Plutonium Fuel," July 1970 4.3, 4.4 0

"Xenon-Induced Spatial Instabilities in Large PWRs," WCAP-3680-20, (EURAEC-1974)

March 1968 4.3 0

"Control Procedures for Xenon-Induced X-Y Instabilities in Large PWR's,"

WCAP-3680-21, (EURAEC-2111) February 1969 4.3 0

"Xenon-Induced Spatial Instabilities in Three-Dimensions," WCAP-3680-22, (EURAEC-2116) September 1969 4.3 0

"Pressurized Water Reactor pH - Reactivity Effect Final Report," WCAP-3698-8, (EURAEC-2074) October 1968 4.3 0

"PUO2 - UO2 Fueled Critical Experiments,"

WCAP-3726-I, July 1967 4.3 0

B/B-UFSAR 1.6-3 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Melting Point of Irradiated UO2,"

WCAP-6065, February 1965 4.2, 4.4 0

"Burnup Physics of Heterogeneous Reactor Lattices," WCAP-6069, June 1965 4.4 0

"LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS,"

WCAP-6073, April 1966 4.3 0

"Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium," WCAP-6086, August 1969 4.3 0

"Subchannel Thermal Analysis of Rod Bundle Cores," WCAP-7015, Revision 1, January 1969 4.4 0

"The PANDA Code," WCAP-7048 (Proprietary) and WCAP-7757 (Non-Proprietary), January 1975 4.3 A

"Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198-L (Proprietary), April 1969 and WCAP-7825 (Non-Proprietary), December 1971 4.3 0

"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7208 (Proprietary), September 1968 and WCAP-7811, (Non-Proprietary), December 1971 4.3 "The TURTLE 24.0 Diffusion Depletion Code,"

WCAP-7213 (Proprietary) and WCAP-7758 (Non-Proprietary), January 1975 4.3, 15.0 15.4 A

"Core Power Capability in Westinghouse PWRs,"

WCAP-7267-L (Proprietary), October 1969 and WCAP-7809 (Non-Proprietary), December 1971 4.3 "Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors,"

WCAP-7306, April 1969 15.4 "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7308, December 1971 4.3 A

B/B-UFSAR 1.6-4 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Application of the THINC Program to PWR Design," WCAP-7359-L (Proprietary), August 1969 and WCAP-7838 (Non-Proprietary),

January 1972 4.4 O

"Seismic Testing of Electrical and Control Equipment," WCAP-7397-L (Proprietary) and WCAP-7817 (Non-Proprietary), December 1971 3.10 O

"Seismic Testing of Electrical and Control Equipment (WCID Process Control Equipment),"

WCAP-7397-L, Supplement 1 (Proprietary) and WCAP-7817, Supplement 1 (Non-Proprietary),

December 1971 3.10 O

"Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," WCAP-7477-L (Proprietary), March 1970 and WCAP-7735 (Non-Proprietary), August 1971 5.2 A

"Radiological Consequences of a Fuel Handling Accident," WCAP-7518-L (Proprietary) and WCAP-7828 (Non-Proprietary), June 1970 15.7 O

"Seismic Vibration Testing with Sine Beats,"

WCAP-7558, October 1972 3.10 O

"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975 15.4 A

"Dynamic Fracture Toughness Properties of Heavy Section A533 Grade B Class 1 Steel Plate," WCAP-7623, December 1970 5.4 O

"Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L (Proprietary) and WCAP-7755 (Non-Proprietary), January 1975 4.4 A

"DNB Tests Results for New Mixing Vane Grids (R)," WCAP-7695-L (Proprietary) and WCAP-7958 (Non-Proprietary) and Addendum, January 1975 4.4 A

B/B-UFSAR 1.6-5 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients,"

WCAP-7706, February 1973 4.6, 7.1 O

"Electric Hydrogen Recombiner for PWR Containments," WCAP-7709-L, Supplements 1 through 7 (Proprietary) and WCAP-7820, Supplements 1 through 7 (Non-Proprietary),

1971 through 1977 3.11, 6.2 A

"A Comprehensive Space-Time Dependent Analysis of Loss of Coolant (SATAN-IV Digital Code),"

WCAP-7750, August 1971 3.6 O

"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, October 1971 15.2 O

"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June 1972 5.2 O

"Behavior of Austenitic Stainless Steel in Post Hypothetical Loss of Coolant Accident Environment," WCAP-7798-L (Proprietary) and WCAP-7803 (Non-Proprietary), January 1972 6.1 O

"Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800, Revision 4-A, April 1975 4.2, 17 A

"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"

WCAP-7806, December 1971 4.3 B

"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7811, December 1971 4.3 O

"Seismic Testing of Electrical and Control Equipment (Low Seismic Plants)," WCAP-7817, Supplements 1-8, December 1971-March 1974 3.10 O

B/B-UFSAR 1.6-6 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Evaluation of Steam Generator Tube, Tubesheet and Divider Plate Under Combined LOCA Plus SSE Conditions," WCAP-7832, December 1973 5.4 A

"Inlet Orificing of Open PWR Cores,"

WCAP-7836, January 1972 4.4 B

"Neutron Shielding Pads," WCAP-7870, May 1972 3.9 A

"LOFTRAN Code Description," WCAP-7907, June 1972 5.2, 15.0 15.1, 15.2, 15.3, 15.4, 15.5, 15.6 A

"FACTRAN - A FORTRAN-IV Code for Thermal Transients in a UO2 Fuel Rod," WCAP-7908, June 1972 15.0, 15.2 15.3, 15.4 A

"MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop PWR System,"

WCAP-7909, June 1972 6.3 O

"Power Peaking Factors," WCAP-7912-L (Proprietary) and WCAP-7912 (Non-Proprietary), January 1975 and Supplement 4.3, 4.4 A

"Damping Values of Nuclear Power Plant Components," WCAP-7921, May 1974 lA, 3.7 A

"Basis for Heatup and Cooldown Limit Curves," WCAP-7924, April 1975 5.3 A

"Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid,"

WCAP-7941-L (Proprietary) and WCAP-7959 (Non-Proprietary), January 1975 4.4 A

"Fuel Assembly Safety Analysis for Combined Seismic and Loss of Coolant Accident, 15x15,"

WCAP-7950, July 1972 3.7 A

"THINC-IV An Improved Program for Thermal and Hydraulic Analysis of Rod Bundle Cores,"

WCAP-7956, June 1973 4.4 A

B/B-UFSAR 1.6-7 REVISION 9 - DECEMBER 2002 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor," WCAP-7964, June 1971 4.3 O

"TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979 (Proprietary) and WCAP-8028 (Non-Proprietary), January 1975 15.0, 15.4 A

"WIT-6 Reactor Transient Analysis Computer Program Description," WCAP-7980, November 1972 15.0, 15.4 A

"Application of Modified Spacer Factor to "L" Grid Typical and Cold Wall Cell DNB,"

WCAP-7988 (Proprietary) and WCAP-8030 (Non-Proprietary), October 1972 4.4 A

"Application of the THINC-IV Program to PWR Design," WCAP-8054 (Proprietary) and WCAP-8195 (Non-Proprietary), October 1973 4.4 A

"Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop,"

WCAP-8082 (Proprietary) and WCAP-8172 (Non-Proprietary), January 1975 3.6 A

"Reactor Coolant Pump Integrity in LOCA,"

WCAP-8163, September 1973 lA, 5.4 O

"Calculational Model for Core Reflooding After a Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974 15.6 A

"Effect of Local Heat Flux Spikes on DNB in Non-Uniform Heated Rod Bundles," WCAP-8174 (Proprietary) and WCAP-8202, (Non-Proprietary), August 1973 4.4 A

"WFLASH, A FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWR," WCAP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision 1 (Non-Proprietary), July 1974 15.6 A

B/B-UFSAR 1.6-8 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Fuel Densification Experimental Results and Model for Reactor Application,"

WCAP-8218 (Proprietary) and WCAP-8219 (Non-Proprietary), March 1975 4.1, 4.2, 4.3, 4.4 A

"Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236 (Proprietary), December 1973 and WCAP-8288 (Non-Proprietary), January 1974 and Addenda 3.7, 4.2 A

"Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March 1974 and WCAP-8288, Addendum 1 (Non-Proprietary), April 1974 3.7 A

"Documentation of Selected Westinghouse Structural Analysis Computer Codes,"

WCAP-8252, Revision 1, July 1977 3.6, 3.9 O

"Hydraulic Flow Test of the 17x17 Fuel Assembly," WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary), February 1974 4.2, 4.4 O

"Effect of 17x17 Fuel Assembly Geometry on DNB," WCAP-8296 (Westinghouse Proprietary) and WCAP-8927 (Non-Proprietary), February 1975 4.4 A

"The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing," WCAP-8298 (Proprietary) and WCAP-8299 (Non-Proprietary),

January 1975 4.4 A

"LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974 15.0, 15.6 AE SATAN-IV Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant,"

WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974 15.0, 15.6 AE

B/B-UFSAR 1.6-9 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests,"

WCAP-8303 (Proprietary) and WCAP-8317 (Non-Proprietary), July 1975 3.9 A

"Control of Delta Ferrite in Austenitic Stainless Steel Weldments," WCAP-8324-A, June 1974 lA, 5.2 A

"Containment Pressure Analysis Code (COCO),"

WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974 15.6 AE "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974 4.3, 4.6, 15.1, 15.2, 15.4, 15.8 O

"Westinghouse ECCS Evaluation Model -

Summary," WCAP-8339, July 1974 6.2, 15.6 AE "Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974 15.6 AE "Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP -8341 (Proprietary) and WCAP-8342 (Non-Proprietary), July 1974 lA(N), 17 A

"Effects of Fuel Densification Power Spikes on Clad Thermal Transients," WCAP-8359, July 1974 4.3 AE "Westinghouse Nuclear Energy Systems Division Quality Assurance Plan," WCAP-8370, Revision 9A, September 1977 1A, 17 A

"Qualification of Westinghouse Seismic Testing Procedure for Electrical Equipment Tested Prior to May 1974," WCAP-8373, August 1974 3.10 O

"Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-Proprietary),

July 1974 4.2 A

B/B-UFSAR 1.6-10 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Power Distribution Control and Load Following Procedures," WCAP-8385 (Proprietary) and WCAP-8403 (Non-Proprietary), September 1974 4.3, 4.4 A

"An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," WCAP-8424, Revision 1, June 1975 15.3 O

"17x17 Drive Line Components Tests - Phase IB, II, III, D-Loop Drop and Deflection,"

WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December 1974 3.9, 15.0 A

"Analysis of Data from the Zion (Unit 1)

THINC Verification Test," WCAP-8453-A (Proprietary), May 1976 and WCAP-8454 (Non-Proprietary), January 1975 4.4 A

"Westinghouse ECCS Evaluation Model -

Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-Proprietary), April 1974 15.6 AE "Incore Power Distribution Determination in Westinghouse Pressurized Water Reactors,"

WCAP-8498, July 1975 4.3 O

"UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations," WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary), April 1975 3.9 A

"Critical Heat Flux Testing of 17x17 Fuel Assembly Geometry with 22 Inch Spacing,"

WCAP-8536 (Proprietary) and WCAP-8537 (Non-Proprietary), May 1975 4.4 A

"Westinghouse ECCS - Four Loop Plant (17x17)

Sensitivity Studies," WCAP-8565 (Proprietary) and WCAP-8566 (Non-Proprietary), July 1975 15.6 A

B/B-UFSAR 1.6-11 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Improved Thermal Design Procedure,"

WCAP-8567-P, July 1975 (Proprietary) and WCAP-8568, July 1975 (Non-Proprietary) 4.4, 15.0 A

"Augmented Startup and Cycle 1 Physics Program Supplement 1," WCAP-8575, June 1976 (Proprietary) and WCAP-8576, June 1976 (Non-Proprietary) and Supplements.

4.3 O

"The Application of Preheat Temperatures After Welding Pressure Vessel Steels,"

WCAP-8577, February 1976 lA A

"Failure Mode and Effects Analysis (FMEA) of the Engineered Safeguard Features Actuation System," WCAP-8584 (Proprietary) and WCAP-8760 (Non-Proprietary), April 1976 4.6 O

"Environmental Qualification of Westinghouse NSSS Class lE Equipment," WCAP-8587, September 1975 lA, 3.1O, 3.11 A

"Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary), November 1975 15.6 A

"Experimental Verification of Wet Fuel Storage Criticality Analyses," WCAP-8682 (Proprietary) and WCAP-8683 (Non-Proprietary), December 1975 4.3 B

"Fuel Rod Bowing," WCAP-8691 (Proprietary) and WCAP-8692 (Non-Proprietary),

December 1975 4.2 O

"Delta Ferrite in Production Austenitic Stainless Steel Weldments," WCAP-8693, January 1976 lA, 5.2 B

"MULTIFLEX - A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708 (Proprietary) and WCAP-8709 (Non-Proprietary), February 1976 3.9 A

B/B-UFSAR 1.6-12 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS Foster, J. P., et al., Westinghouse Improved Performance Analysis and Design Model (PAD 4.0), WCAP-15063-P-A, Revision 1 with Errata, July 2000.

4.2 A

"New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, July 1976 (Proprietary) and WCAP-8763, July 1976 (Non-Proprietary) 4.4 A

"Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, Revision 2, October 1978 1.5, 4.2, 4.3 B

"Verification of Neutron Pad and 17x17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant," WCAP-8780, May 1976 3.9 B

"Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations,"

WCAP-8785, October 1976 4.2 "Hybrid B4C Absorber Control Rod Evaluation Report," WCAP-8846, October 1977 4.2, 15.0 15.3 A

"Westinghouse ECCS - Four Loop Plant (17x17)

Sensitivity Studies with Upper Head Fluid Temperature at Thot," WCAP-8865, May 1977 15.6 A

"7300 Series Process Control System Noise Tests," WCAP-8892-A, April 1977 7.1 A

"Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8963 (Proprietary), November 1976 and WCAP-8964 (Non-Proprietary), August 1977 4.2 A

"Westinghouse Emergency Core Cooling System Small Break October 1975 Model," WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary),

April 1977 15.6 A

B/B-UFSAR 1.6-13 REVISION 1 - DECEMBER 1989 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Failure Mode and Effects Analysis of the Solid State Full Length Rod Control System,"

WCAP-8976, September 1977 4.6 O

"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"

WCAP-9000-L, Revision 1 (Proprietary), July 1969 and WCAP-7806 (Non-Proprietary), December 1971.

4.3 "Axial Power Distribution Monitoring Using Four-Section Ex-Core Detectors," WCAP-9105 (Proprietary) and WCAP-9106 (Non-Proprietary),

July 1977 4.3 A

"Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCAs During Operation with One Loop Out of Service for Plants Without Loop Isolation Valves,"

WCAP-9166 (Proprietary) and WCAP-9167 (Non-Proprietary), February 1978 15.6 O

"Westinghouse Emergency Core Cooling System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9150 (Non-Proprietary), September 1977 15.6 O

"Properties of Fuel and Core Component Materials," WCAP-9179 (Proprietary), September 1977 and WCAP-9224 (Non-Proprietary) 4.2 O

"Westinghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220 (Proprietary Version), WCAP-9221 (Non-Proprietary Version), February 1978 15.6 A

"Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly," WCAP-9401 (Proprietary) and WCAP-9402 (Non-Proprietary), March 1979 4.1, 4.2, 4.4 A

"PALADON - Westinghouse Nodal Computer Code,"

WCAP-9485 (Proprietary) and WCAP-9486 (Non-Proprietary) December 1978 4.3 A

B/B-UFSAR 1.6-14 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Reference Core Report 17x17 Optimized Fuel Assembly," WCAP-9500 (Non Proprietary),

July 1979 4, 15 A

"RELAP5/MOD2-B&W - An Advanced Computer Code for Light Water Reactor LOCA and non-LOCA Transient Analysis" BAW-10164, Revision 3 (non-proprietary), October 1996 15 A

"CONTEMPT - Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident,", BAW-10095A, Revision 1, April 1978 6

O Beacon Core Monitoring and Operations Support System, WCAP-12472 (Proprietary Class 2),

August 1994 4.3, 4.4, 7.7 A

Relaxation of Constant Axial Offset Control, FQ Surveillance Technical Specification, WCAP-10216-P-A, Revision 1A (Proprietary Class 2), February 1994 4.3, 4.4 A

VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, WCAP-14565-P-A (Proprietary) / WCAP-15306-NP-A (Non-Proprietary), October 1999 4.4, 15.0 A

Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications, WCAP-14565-P-A Addendum 2-P-A (Proprietary) / WCAP-15306-NP-A Addendum 2-NP-A (Non-Proprietary), April 2008 4.4 A

SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, WCAP-10698-P-A (Proprietary) and WCAP-10750-A (Non-Proprietary), August 1987 15.0 A

B/B-UFSAR 1.7-1 REVISION 9 - DECEMBER 2002 1.7 DRAWINGS The drawings cited in each UFSAR Chapter 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. References on the figures contained in the UFSAR to ComEd, CECo, and Commonwealth Edison will be revised to reflect the change in facility ownership to Exelon Generation Company when other changes to that figure are needed.

1.7.1 Electrical, Instrumentation, and Control Drawings Subsection 1.7.1 of the FSAR provides a list of electrical, instrumentation, and control drawings that were provided to the NRC during the initial licensing phase.

1.7.2 Drawings for Independent Structural Review Subsection 1.7.2 of the FSAR provides a list of the structural, architectural, mechanical loading and electrical loading drawings that were provided to the NRC to enable them to perform the Project Structural Review and the Independent Structural Review during the licensing phase.

B/B-UFSAR 1.7-2 REVISION 9 - DECEMBER 2002 TABLE 1.7-1 This Table has been intentionally deleted.

B/B-UFSAR REVISION 9 - DECEMBER 2002 Pages 1.7-3 through 1.7-17 have been intentionally deleted.

B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.1-1 through 1.1-3 have been deleted intentionally.

B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.2-1 through 1.2-17 have been deleted intentionally.

B/B-UFSAR 1.0-i REVISION 5 - DECEMBER 1994 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT 1.1-1

1.1 INTRODUCTION

1.1-1 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Site and Environment 1.2-1 1.2.2 Nuclear Steam Supply System 1.2-1 1.2.3 Engineered Safety Features 1.2-2 1.2.4 Emergency Core Cooling System 1.2-3 1.2.5 Control and Instrumentation 1.2-3 1.2.6 Electrical System 1.2-4 1.2.7 Turbine and Auxiliaries 1.2-4 1.2.8 Fuel Handling System 1.2-5 1.2.9 Radioactive Waste Management System 1.2-5 1.2.10 Features of Special Interest 1.2-5 1.2.11 Structures 1.2-6 1.3 COMPARISON TABLES 1.3-1 1.3.1 Comparisons with Similar Facility Designs 1.3-1 1.3.2 Comparison of Final and Preliminary Information 1.3-2 1.3.3 References 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Licensee 1.4-1 1.4.2 Architect-Engineer 1.4-1 1.4.3 Reactor Designer 1.4-1 1.4.4 Constructor 1.4-2 1.4.5 Consultants and Service Organization 1.4-2 1.4.5.1 Security Systems - ETA 1.4-2 1.4.5.2 Dames & Moore 1.4-3 1.4.5.3 HARZA Engineering 1.4-3 1.4.5.4 Murray and Trettel, Inc.

1.4-3 1.4.5.5 Shirmer Engineering Corporation 1.4-3 1.4.5.6 Hyla S. Napadensky 1.4-4 1.4.5.7 NALCO Chemical Company 1.4-4 1.4.5.8 Westinghouse Environmental Systems Department (WESD) 1.4-4 1.4.5.9 Illinois Natural History Survey (INHS) 1.4-5 1.4.5.10 NUS Corporation 1.4-5 1.4.5.11 Eberline Instrument Corporation (EIC) 1.4-5

B/B-UFSAR 1.0-ii REVISION 9 - DECEMBER 2002 TABLE OF CONTENTS (Cont'd)

PAGE 1.4.5.12 Meteorology Research, Inc. (MRI) 1.4-5 1.4.5.13 Illinois State Museum (ISM) 1.4-6 1.4.5.14 Equitable Environmental Health, Inc. (EEH) 1.4-6 1.4.5.15 Espey, Huston & Associates, Inc. (EH & A) 1.4-6 1.4.5.16 University of Wisconsin-Milwaukee (UWM) 1.4-7 1.4.5.17 Aero-Metric Engineering, Inc. (AME) 1.4-7 1.4.5.18 Iowa Institute of Hydraulic Research 1.4-7 1.4.5.19 Babcock and Wilcox International (B&W) 1.4-8 1.4.5.20 Framatome Technologies, Incorporated (FTI) 1.4-8 1.4.5.21 Stone & Webster Engineers and Constructors, Inc, (S&W) 1.4-8 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Programs Required For Plant Operation 1.5-1 1.5.1.1 Core Stability Evaluation 1.5-1 1.5.2 Other Programs Not Required For Plant Operation 1.5-1 1.5.2.1 Fuel Development Program For Operation at High Power Densities 1.5-2 1.5.2.2 Blowdown Forces Program 1.5-2 1.5.2.3 Blowdown Heat Transfer Testing 1.5-2 1.5.3 References 1.5-4 1.6 MATERIAL INCORPORATED BY REFERENCES 1.6-1 1.7 DRAWINGS 1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Drawings for Independent Structural Review 1.7-1

B/B-UFSAR 1.0-iii REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF TABLES NUMBER TITLE PAGE 1.3-1 Plants Using Three-Buttress Containment Design 1.3-3 1.4-1 Exelon Generation Company's Nuclear Power Plants in Service or Under Construction 1.4-9 1.4-2 Nuclear Power Plants Completed or Currently Under Design by Sargent & Lundy 1.4-10 1.4-3 Westinghouse Pressurized Water Reactor Nuclear Power Plants 1.4-11 1.5-1 Blowdown Heat Transfer Phase I Test Parameters 1.5-5 1.5-2 Blowdown Heat Transfer Phase II Test Parameters 1.5-6 1.6-1 Topical Reports Incorporated by Reference 1.6-2 1.7-1 Deleted 1.7-2

B/B-UFSAR 1.0-iv REVISION 9 - DECEMBER 2002 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 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.

DRAWINGS*

SUBJECT M-1 General Site Plan Units 1 & 2 M-2 Property Development Units 1 & 2 M-5 General Arrangement Roof Plan Units 1 & 2 M-6 General Arrangement Main Floor At El. 451-0 Units 1

& 2 M-7 General Arrangement Mezzanine Floor At El. 426-0 Units 1 & 2 M-8 General Arrangement Grade Floor At El. 401-0 Units 1 & 2 M-9 General Arrangement Floor Plan At El. 383-0 Units 1

& 2 M-10 General Arrangement Basement Floor At El. 364-0 Units 1 & 2 M-11 General Arrangement Floor Plan At El. 346-0 Units 1

& 2 M-12 General Arrangement Radwaste/Service Building Units 1

& 2 M-13 General Arrangement Fuel Handling Building Units 1 &

2 M-14 General Arrangement Section A-A Units 1 & 2 M-15 General Arrangement Section B-B Units 1 & 2 M-16 General Arrangement Section C-C and D-D Units 1 &

2 M-17 General Arrangement Section E-E Units 1 & 2 M-18 General Arrangement Section F-F Units 1 & 2 M-19 General Arrangement Lake Screen House Units 1 & 2 (Braidwood)

M-20 General Arrangement River Screen House Units 1 & 2 M-22 General Arrangement Miscellaneous Plans Units 1 & 2 M-34 P&ID Index and Symbols Units 1 & 2

B/B-UFSAR 1.1-1 REVISION 15 - DECEMBER 2014 CHAPTER 1.0 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

The Nuclear Regulatory Commission approved the transfer of the facility licenses from Commonwealth Edison (ComEd) Company to Exelon Generation Company, LLC (EGC) on August 3, 2000 (Reference 1). References in the Updated Final Safety Analysis Report (UFSAR) to ComEd, CECo, and Commonwealth Edison have been retained, as appropriate, instead of being changed to EGC to properly preserve the historical context.

This UFSAR is submitted by Exelon Generation Company for nuclear power plants at Byron, Illinois and at Braidwood, Illinois (Drawings M-1 and M-2) in accordance with the requirements of 10 CFR 50.71(e). Each power plant consists of two units having nearly identical nuclear steam supply systems (NSSS) and turbine generators. The main exception is that the original Unit 1 steam generators were replaced by steam generators of a different design. The power plants at the two sites are as nearly identical as site characteristics permit. The bulk of this UFSAR applies to the standardized, non-site-related aspects of the power plants. Sections which describe features specific to the sites are repeated for each site and the applicable station name appears at the top of these pages. Every effort has been made in the preparation of this document to conform to the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants", Revision 2, September 1975. The guidance provided in Nuclear Energy Institute (NEI) 98-03, Guidelines for Updating Final Safety Analysis Reports, Revision 1, June 1999, as endorsed by NRC Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10CFR50.71(e), Revision 0, September 1999, is used to comply with the provisions of 10CFR50.71(e).

Each nuclear power plant consists of two nearly identical generating units, and two pressurized water reactor (PWR) (NSSS) and turbine-generator furnished by Westinghouse Electric Corporation (Westinghouse) similar in design concept to several projects recently licensed or currently under review by the NRC (see Section 1.3). Unit 1 contains steam generators supplied by B&W and Unit 2 contains steam generators supplied by Westinghouse.

Westinghouse Electric Corporation, Sargent &

Lundy, and the Commonwealth Edison Company jointly participated in the original design and construction of each unit. The plant is operated by Exelon Generation Company. Sargent & Lundy (S&L) is the architect-engineer for both stations.

Each nuclear steam supply system (NSSS) has been evaluated at a power output of 3672 MWt for the Measurement Uncertainty Recapture (MUR) Power Uprate. The warranted gross and approximate net electrical outputs for the MUR are 1268 MWe and 1241 MWe for Unit 1 and Unit 2, respectively. Safety analyses are evaluated at an NSSS power level of 3672 MWt and a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level of 3648 MWt.

B/B-UFSAR 1.1-1a REVISION 15 - DECEMBER 2014 Specifically, the containment and engineered safety features (ESF) are designed and evaluated for operation at a core thermal power level of 3658 MWt. Accidents (such as loss-of-coolant, steamline break, and other postulated accidents having offsite dose consequences) are also analyzed at a core thermal power level of 3658 MWt. DNB analyses are evaluated at a core thermal power level 3648 MWt.

B/B-UFSAR 1.1-2 REVISION 9 - DECEMBER 2002 The reactor containments are of post-tensioned concrete construction with a carbon steel liner. Sufficient free volume is provided to contain the energy released in a major accident without need for "pressure suppression" devices. Sargent & Lundy is responsible for containment design.

Byron Station is located in north central Illinois, near the town of Byron and near the Rock River (Drawing M-1). Cooling for the plant is provided by two natural draft cooling towers for non-essential service cooling, and by mechanical draft cooling towers for essential cooling. The fuel loading dates for the two units were November 1984 and November 1986 for Units 1 and 2, respectively. The corresponding dates for commercia1 operation were September 1985 and August 1987.

The Braidwood Station is located in northeastern Illinois, near the town of Braidwood and near the Kankakee River (Drawing M-1).

Cooling for the plant is provided by a large man-made cooling pond of approximately 2500 acres constructed over a previously strip-mined area. Essential service cooling is provided by a 99-acre auxiliary cooling pond which is integral with the main pond. The fuel loading dates for the two units were October 1986 and December 1987 for Units 1 and 2, respectively. The corresponding dates for commercial operation were July 1988 and October 1988.

The standard symbols used on piping and instrument diagrams and other figures in this UFSAR are shown in Drawing M-34.

1.2 REFERENCES

1.

NRC letter, "Braidwood, Byron, Dresden, LaSalle, Quad Cities, and Zion - Orders Approving Transfer of Licenses From Commonwealth Edison Company To Exelon Generation Company, LLC, and Approving Conforming Amendments," dated August 3, 2000

B/B-UFSAR 1.2-1 REVISION 15 - DECEMBER 2014 1.2 GENERAL PLANT DESCRIPTION 1.2.1 Site and Environment The characteristics of the sites and their environs have been investigated to establish bases for determining criteria for storm, flood, and earthquake protection and to evaluate the validity of calculational techniques for the control of routine and accidental releases of radioactive liquids and gases to the environment. Field programs to investigate geology and seismology are completed. Preoperational meteorological programs to provide onsite observations of wind speed and direction have continued since the spring of 1973 at Byron and since the fall of 1973 for Braidwood. Radiological studies of the site environs were initiated at least 18 months prior to commercial operation, with the objective of establishing background radiation levels.

The geography, demography, meteorology, hydrology, geology, and seismology of the two plant sites are discussed in detail in Chapter 2.0.

1.2.2 Nuclear Steam Supply System The nuclear steam supply system (NSSS) consists of a Westinghouse pressurized water reactor and supporting auxiliary systems.

Performance at the calculated steam flow of the NSSS at MUR conditions based on zero percent makeup is as follows:

a.

thermal output of NSSS (MWt) - 3659; b.

thermal output of reactor core (MWt) -3645; c.

steam flow from NSSS (lb/hr) - 16,347,514 for Unit 1/16,280,677 for Unit 2; d.

steam pressure at a steam generator outlet (psia) -

1020.8 for Unit 1 and 902 for Unit 2; e.

maximum moisture content (%) - 0.25%; and f.

feedwater temperature at steam generator inlet (F) -

446.5 for Unit 1 and 447.5 for Unit 2.

The NSSS consists of a reactor and closed reactor coolant loops connected in parallel to the reactor vessel, each loop containing a reactor coolant pump and a steam generator. The NSSS also contains an electrically heated pressurizer and certain auxiliary systems.

B/B-UFSAR 1.2-1a REVISION 7 - DECEMBER 1998 High pressure reactor coolant circulates through the reactor core to remove the heat generated by the nuclear reaction. The heated reactor coolant flows from the reactor vessel to the steam generators (via reactor coolant loop piping). The coolant gives up its heat to the feedwater in the steam generator to generate steam for the turbine generator. The cycle is completed when the reactor coolant is pumped back to the reactor vessel. The entire reactor coolant system is composed of leaktight components to contain the reactor coolant to the system.

B/B-UFSAR 1.2-2 REVISION 11 - DECEMBER 2006 The core is a multiregion type. All fuel assemblies are mechanically identical, although the fuel enrichment is not the same in all assemblies. In a typical initial core loading, three fuel enrichments are used in mechanically identical assemblies.

Fuel assemblies with the highest enrichments are placed in the core periphery, or outer region, and the two groups of lower enrichment fuel assemblies are arranged in a selected pattern in the central region. In subsequent refuelings, one third of the fuel is discharged and fresh fuel is loaded into the outer region of the core. The remaining fuel is arranged in the central two-thirds of the core in such a manner as to achieve optimum power distribution.

Rod cluster control assemblies are used for reactor control and consist of clusters of cylindrical absorber rods. The absorber rods move within guide tubes in certain fuel assemblies. Above the core, each cluster of absorber rods is attached to a spider connector and drive shaft, which is raised and lowered by a drive mechanism mounted on the reactor vessel head. The insertion of the rod cluster control assembly for a reactor trip is by gravity.

The reactor coolant pumps are Westinghouse vertical, single-stage, centrifugal pumps of the shaft-seal type.

The steam generators are B&W vertical U-tube units for Unit 1 and Westinghouse vertical U-tube units for Unit 2. All steam generators contain Inconel tubes. Integral moisture separation equipment reduces the moisture content of the steam.

The reactor coolant piping and all of the pressure-containing surfaces in contact with reactor water are stainless steel. The steam generator tubes and fuel cladding are Inconel and Zircaloy/ZIRLO, respectively. Reactor core internals, including control rod drive shafts, are primarily stainless steel.

An electrically heated pressurizer connected to one reactor coolant loop maintains reactor coolant system pressure during normal operation, limits pressure variations during plant load transients, and keeps system pressure within design limits during abnormal conditions.

Auxiliary system components are provided to charge makeup water into the reactor coolant system, purify reactor coolant, provide chemicals for corrosion inhibition and reactivity control, cool system components, remove decay heat, and provide for emergency safety injection.

1.2.3 Engineered Safety Features The engineered safety features provided for this plant have sufficient redundancy of components and power sources such that

B/B-UFSAR 1.2-3 REVISION 12 - DECEMBER 2008 under the conditions of a loss-of-coolant accident they can maintain the containment integrity and limit personnel exposure to less than 10 CFR 50.67 limits. The engineered safety features incorporated in the design of this plant and the functions they serve are summarized in the following.

a.

The emergency core cooling system injects borated water into the reactor coolant system if coolant is lost. This system limits damage to the core and limits the fission product contamination released into the containment following a postulated loss-of-coolant accident (LOCA).

b.

A steel lined, concrete containment vessel consists of a post-tensioned concrete cylindrical wall and shallow dome, and a conventionally reinforced concrete base. The containment forms a virtually leaktight barrier to prevent the escape of radioactivity.

c.

Reactor containment fan coolers reduce containment temperature and pressure following a postulated loss-of-coolant accident.

d.

A containment spray system is used to reduce containment pressure and to remove iodine and particulate fission products from the containment atmosphere in the event of a loss-of-coolant accident.

e.

The auxiliary feedwater system provides for heat removal from the reactor coolant system by providing makeup water to the steam generator under a variety of postulated conditions.

f.

A combustible gas control system is provided to ensure that the containment atmosphere is mixed following a loss-of-coolant accident. A mixed containment atmosphere prevents local accumulation of combustible or detonable gases that could threaten containment integrity or equipment operating in a local compartment.

1.2.4 Emergency Core Cooling System The emergency core cooling system (ECCS), with passive and active subsystems, is designed to inject borated water into the reactor coolant system (RCS) following a LOCA. This will provide cooling to limit core damage, metal-water reactions, and fission-product release. The ECCS provides long-term postaccident cooling of the core by drawing borated water from the containment sump.

1.2.5 Control and Instrumentation The reactor is controlled by a variety of reactivity coefficients (temperature, pressure, doppler) by control rod cluster motion which is required for load follow transients and for startup and shutdown, and by a soluble neutron absorber, i.e., boron in the

B/B-UFSAR 1.2-4 REVISION 15 - DECEMBER 2014 form of boric acid which is adjusted in concentration to compensate for such effects as fuel consumption and accumulation of fission products.

1.2.6 Electrical System Each unit's main generator is an 1800-rpm, 3-phase, 60-cycle, hydrogen-innercooled unit with water-cooled stator windings and is rated at 1361 MVA at 75 psig gas pressure. Field excitation is provided by a direct shaft-driven brushless excitation system. Two one-half size main step-up transformers deliver power to the 345-kV switchyard.

The station's auxiliary power system consists of system and unit auxiliary transformers; 6900-V, 4160-V, and 480-V switchgear; 480-V motor control centers; 120-Vac instrument buses; and 250-Vdc and 125-Vdc buses.

Two diesel generators are provided for each unit and are available as onsite sources of power (in the event of complete loss of normal a-c power) for operating essential safeguard features. Each diesel generator is capable of supplying required electrical loads for a simultaneous LOCA and loss of offsite power to any one unit.

1.2.7 Turbine and Auxiliaries The turbine for each unit is a four-casing, tandem-compound, six-flow exhaust, 1800-rpm unit with 40-inch last-row blades. There are two combination moisture-separator/steam-reheater assemblies per unit. The turbine-generator for Units 1 have a MUR rating of 1268 MWe gross at 16,347,514 lb/hr steam flow with inlet steam conditions of 1001 psia, 0.36% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. The turbine-generators for Units 2 have a MUR rating of 1241 MWe gross at 16,280,677 lb/hr steam flow with inlet steam conditions of 882 psi, 0.34% moisture, exhausting at 3.5 in. Hg abs, at zero percent makeup. There are seven stages of feedwater heating for all units.

The turbine is equipped with a redundant fault tolerant Westinghouse Ovation based distributed control system. All control algorithms and processes within the turbine control system are redundant and configured to allow unrestricted turbine operation. This system utilizes a fire-resistant hydraulic fluid to control throttle and governor valve positioning.

B/B-UFSAR 1.2-4a REVISION 11 - DECEMBER 2006 The condenser is of the single-pass deaerating type. There are three parallel strings of feedwater heaters that utilize extraction steam from the low pressure turbines, two parallel strings of feedwater heaters that utilize extraction and exhaust steam from the high pressure turbine, four one-third-sized feedwater condensate and condensate booster pumps, and three one-half-sized feedwater and heater drain pumps. Heater drains from the three highest-pressure feedwater heaters are pumped into the feedwater system; drains from the four lowest-pressure heaters are cascaded to the condenser.

B/B-UFSAR 1.2-5 REVISION 14 - DECEMBER 2012 1.2.8 Fuel Handling System The reactor is refueled with equipment which handles the spent fuel under water from the entire time from leaving the reactor vessel until it is secured in a cask for shipment. Underwater transfer of spent fuel provides a transparent radiation shield and a reliable coolant for decay heat removal.

Fuel handling is performed in the refueling cavity which is flooded for refueling, and the fuel storage pool which is in the fuel building. The two areas are connected by a fuel transfer system which carries the fuel through an opening in the reactor containment.

Spent fuel is removed from the reactor vessel by a refueling machine, placed on the fuel transfer cart conveyor and transferred to the spent fuel storage pool. The fuel is removed from the transfer cart and placed into storage racks. After a suitable decay period, the fuel may be removed from storage and loaded into a shipping cask for removal from the plant.

Refer to Section 9.1.2.3.11 for a description of spent fuel storage and handling using Dry Cask Storage (DCS) system and the Independent Spent Fuel Storage Installation (ISFSI).

All important pumps, piping, and equipment are replicated and capable of being supplied from one of two independent ESF divisions.

1.2.9 Radioactive Waste Management System The radioactive waste system provides equipment necessary to collect, process, and prepare for the disposal of radioactive liquid, gaseous, and solid wastes produced as a result of reactor operation or to transfer the wastes to a vendor-supplied radwaste system.

After collection, depending on chemical composition, liquid wastes may be demineralized and/or filtered. The treated water is discharged at concentrations within the limits of 10 CFR 20.

Sludges and spent demineralizer resins are processed by a vendor-supplied radwaste system for ultimate disposal in an authorized location.

Gaseous wastes are collected from the waste gas header.

Discharge of the gaseous wastes to the environment is controlled to ensure that the offsite dose is as low as reasonably achievable (ALARA).

1.2.10 Features of Special Interest The fundamental concept for the design and construction of the Byron/Braidwood Stations is one of commonality and duplication to the maximum extent permitted by site characteristics. For those features not dictated specifically by site characteristics, identical designs have been employed for the two stations. The concept has been extended to the point where the limiting (i.e.,

B/B-UFSAR 1.2-6 REVISION 9 - DECEMBER 2002 worst case) parameters of the sites are considered in the common design. An example of this is the use of the most restrictive site's seismic building response spectra for the design of systems and components in both plants.

Common plans, drawings, and specifications have been issued for construction at the two sites. Design and construction management for both sites have been conducted by the same major organizations, using the same quality assurance and project management programs. This approach embraces the concept of standardization in nuclear power plant design and construction.

1.2.11 Structures The major structures include a separate and independent containment for each reactor, a common auxiliary building, a common turbine building, a common solid radwaste storage, and administration and service building. General layouts of the plant and interior components' arrangements are shown on Drawings M-5 through M-18 and M-20 and M-22 (Byron), and Drawings M-5 through M-20 and M-22 (Braidwood).

For purposes of design and analysis, structures are designated by Safety Category according to their relation to plant safety. The Safety Category definitions are as follows:

a.

Safety Category I - Those structures important to safety that must be designed to remain functional in the event of the safe shutdown earthquake (SSE) and other design-basis events (including tornado, probable maximum flood, operating basis earthquake (OBE), missile impact, or accident internal to the plant) are designated as Safety Category I.

b.

Safety Category II - Those structures which are not designated as Safety Category I are designated as Safety Category II.

The design criteria and analysis methods for these structures are discussed in Chapter 3.0.

B/B-UFSAR 1.3-1 REVISION 8 - DECEMBER 2000 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs The design is conceptually similar to Exelon Generation Company's Zion Station. Differences in the design of the two plants have been allowed only (1) when dictated by the site characteristics, (2) when the change would result in significant safety improvement, simplification of construction or operation procedures, or cost savings; or (3) as required to comply with appropriate codes and standards, NRC criteria, regulatory guides, and regulations.

The nuclear steam supply system is similar to that of the Zion Station but has a slightly higher power rating. The reactor containments are of the same materials and size as those at the Zion Station, but each has only three buttresses, rather than six as used at Zion. The number of post-tensioning tendons is reduced, and the number of wires per tendon increased, from that used at Zion. The reduced number of buttresses allows for greater separation of penetration areas for redundant safety-related systems. Several plants on which this buttress design has been used are listed in Table 1.3-1.

The polar cranes in the reactor containment are mounted on the containment wall, rather than on the missile barrier as at Zion.

This allows use of a greater area for component laydown in the containment.

Two 100%-capacity containment spray systems are used, rather than the three systems used at Zion. Four containment fan coolers are used, rather than the five used at Zion. The emergency diesel-generator systems for each unit are entirely independent and use two 5500-kW diesel generators per unit. The arrangement of equipment in the common auxiliary building allows greater physical separation of redundant systems and their piping and cables than was possible at Zion.

The Byron Station uses natural draft cooling towers for heat rejection. Zion utilizes once-through cooling. Mechanical draft cooling towers are provided for essential service cooling at Byron.

The Braidwood Station uses a large man-made cooling pond for heat rejection. An auxiliary cooling pond, integral with the main pond, is provided for essential service cooling.

Table 1.3-2 of the FSAR provided the design comparison of the Byron/Braidwood nuclear steam supply system with Comanche Peak, Indian Point 2, South Texas, Sun Desert, W. B. McGuire Nuclear Station, Trojan Nuclear Power Plant, SNUPPS, and the Watts Bar Application. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.

B/B-UFSAR 1.3-2 REVISION 8 - DECEMBER 2000 1.3.2 Comparison of Final and Preliminary Information The Byron/Braidwood Power Plant design was subject to continuing review throughout the construction of the stations. The experience gained at Zion Station and other PWRs was used to enhance equipment reliability and performance. Current design technology was used to upgrade earlier plant design methods.

No significant design changes have been made to the Byron Station or the Braidwood Station which have not been previously reported by amendment to the PSAR, except for the inclusion of 17 x 17 optimized fuel. Table 1.3-3 of the FSAR listed those significant changes reported since the issuance of the Byron and Braidwood Stations Construction Permits. This information was current at the time the Byron Unit 1 operating license was granted and has not been included in the UFSAR.

Other changes included the removal of the part length control rods (they are not needed to control Xenon induced axial instabilities), the enlargement of spent fuel capacity, the use of more corrosion-resistant materials in the steam generators and moisture steam separators, improved equipment packaging to do a reactor refueling in a shorter time period, an upgraded design for the reactor coolant pump seals, and replacement steam generators for Unit 1. These concepts are described in later chapters.

1.3.3 References

1. Exelon Generation Company, "Byron/Braidwood Stations Fire Protection Report in Response to Appendix A of BTP APCSB 9.5-1,"

(current amendment).

B/B-UFSAR 1.3-3 TABLE 1.3-1 PLANTS USING THREE-BUTTRESS CONTAINMENT DESIGN PLANT/UTILITY DATE OF OPERATION Arkansas Nuclear One Arkansas Power & Light Co.

5-21-74 Millstone-2 Northeast Utilities 8-1-75 Rancho Seco Sacramento Municipal Utility District 8-16-74 Trojan Portland General Electric Co.

11-21-75 J.M. Farley-1 Alabama Power Co.

6-25-77

B/B-UFSAR 1.4-1 REVISION 8 - DECEMBER 2000 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 Licensee Exelon Generation Company is the Licensee for the Byron Station, Units 1 and 2, which is located in Rockvale Township, Ogle County, approximately 4 miles south of Byron, Illinois, and for Units 1 and 2 of the Braidwood Station, which is located in Reed Township, Will County, approximately 6 miles southwest of Wilmington, Illinois. The Licensee is responsible for the design, construction, and operation of the nuclear power plants.

Commonwealth Edison supplies electrical service to an area of 13,000 square miles with a population of approximately 8 million persons, located primarily in the northern third of Illinois.

Dresden 1, Commonwealth Edison's first nuclear generating station, went into commercial service during August 1960, and has produced more than 10 billion kWh. Additional nuclear units in service or under construction are listed in Table 1.4-1.

1.4.2 Architect-Enqineer For the work covered by this application, Sargent & Lundy (S&L) has been retained as the design consultants. The Licensee has employed Sargent & Lundy for power plant design work for over 80 years.

Sargent & Lundy is an independent consulting engineering organization founded in Chicago, in 1891. For over three-quarters of a century, the firm has specialized exclusively in the design of generation, transmission, distribution, and utilization of steam and electric power and related facilities.

The firm has provided the complete engineering services for more than 600 turbine-generator units with a total capacity of 53,000,000 kW. Of this total, some 9,800,000 kW is in nuclear generating capacity. Table 1.4-2 lists the nuclear plants completed by or currently under design by Sargent & Lundy.

1.4.3 Reactor Designer Westinghouse has designed, developed, and manufactured nuclear power facilities since the 1950s, beginning with the world's first large central station nuclear power plant (Shippingport),

which started producing power in 1957. Completed or contracted

B/B-UFSAR 1.4-2 REVISION 5 - DECEMBER 1994 commercial nuclear capacity totals were in excess of 98,000 MWe. Westinghouse pioneered new nuclear design concepts, such as chemical shim control of reactivity and the rod cluster control concept, throughout the last two decades. Westinghouse manufacturing facilities include the largest commercial nuclear fuel fabrication facility in the world and the world's most modern heat transfer equipment production facility, as well as other facilities producing nuclear steam supply system (NSSS) components. Table 1.4-3 lists all Westinghouse pressurized water reactor (PWR) plants to date, including those plants under construction or on order at the time of the Byron/Braidwood application.

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute have contracted with Westinghouse for research into NSSS-related activities. Westinghouse experience was also utilized by the NRC and Metropolitan Edison immediately following the Three Mile Island Unit 2 accident and the corporation continues to participate with the Westinghouse Owner's Group of utilities in addressing the NRC action plan and other operations improvements.

1.4.4 Constructor Construction coordination of all activities at the site was under the supervision of the Commonwealth Edison's Station Construction Department. The department exercises site managerial functions as discussed in Chapter 17.0 of the UFSAR.

The Station Construction Department was the constructor for Zion Station. This department has coordinated the construction activities for almost all of Commonwealth Edison's existing power plants. It was also the construction coordinator for La Salle County Station.

1.4.5 Consultants and Service Organization 1.4.5.1 Security System - ETA The design of the physical security system and the administrative controls was performed by ETA, Inc.

ETA personnel have had varied and in-depth experience in the design, safety analysis, and environmental review of nuclear power plants and related facilities as well as in the management and organization of security systems. They are very familiar with the details of the current generation of light water reactors and, in particular, those critical areas and components of the plants which might be the most vulnerable to sabotage.

They are also familiar with the current regulations and guidelines of the NRC that define the required performance and objectives of a security system for licensed activities.

B/B-UFSAR 1.4-3 1.4.5.2 Dames & Moore The independent consulting firm of Dames & Moore was employed to conduct studies relating to the geology, seismology, and groundwater hydrology at both sites. The firm also conducted preconstruction baseline studies, including wildlife surveys as well as soil and vegetation analyses.

Having performed environmental studies for approximately 30 nuclear power plant sites, Dames & Moore is a recognized authority in the field of environmental engineering of nuclear power plants.

1.4.5.3 HARZA Engineering HARZA was employed in the design of the water treatment facilities at both stations.

HARZA has been involved with a variety of technical studies for at least ten nuclear power stations. Among these studies have been facility design, review of design and structure, hydrology, and groundwater. In addition, HARZA Engineering has designed some of the largest hydroelectric projects in the world, including major concrete structures and earthfilled dams.

1.4.5.4 Murray and Trettel, Inc.

Murray and Trettel (M&T) is an environmental consulting firm which, since 1960, has provided significant meteorological input to both preoperational and operational phases of meteorological programs for nuclear power stations. M&T has also provided meteorological input to a wide variety of air pollution and environmental problems as well as allied control technique programs.

Murray & Trettel provided meteorological data for both stations by implementation of an onsite measurement program incorporating a tower for elevation measurements.

1.4.5.5 Shirmer Engineering Corporation Shirmer Engineering is a firm of consulting fire protection engineers. The firm has done work on 17 Department of Energy nuclear fuel production and laboratory facilities, as well as for numerous nuclear power stations for Sargent & Lundy.

Shirmer Engineering has also performed services for many fossil units.

Shirmer Engineering provided evaluation of the fire protection systems at both stations and assisted in the preparation of the Byron/Braidwood Fire Protection Report.

B/B-UFSAR 1.4-4 1.4.5.6 Hyla S. Napadensky Ms. Napadensky was retained to help evaluate the probability of an accidental explosion occurring on a train carrying explosives in the vicinity of the Braidwood Station.

Ms. Napadensky is the Manager of Fire Safety Research at the IIT Research Institute of the Illinois Institute of Technology.

Ms. Napadensky has directed analytical and experimental research in the areas of explosion effects, hazards and risk analysis, safety of chemical systems, explosives and propellant sensitivity, and initiation mechanisms during her 17 years with IIT Research Institute.

1.4.5.7 NALCO Chemical Company The NALCO Chemical Company (formerly Industrial Bio-Test, Inc.)

consisted of two divisions, Industrial Bio-Test Laboratories, and NALCO Environmental Sciences, which conduct studies relating to toxicology and ecological sciences, respectively. The Environmental Science Division includes seven subdivisions: (1) aquatic biology, (2) fisheries and field operations, (3) water and wastewater chemistry, (4) radiochemistry, (5) air sciences and data processing, (6) land and plant sciences, and (7) environmental physiology.

As a technical consultant on the Braidwood project, the NALCO Chemical Company provided a clam bed mapping survey in the area of the station's intake and discharge structures located on the Kankakee River.

1.4.5.8 Westinghouse Environmental Systems Department (WESD)

WESD, established as a department of the Westinghouse Power Systems Company in 1969, consisted of environmental scientists and engineers experienced in the areas of aquatic and terrestrial biology and ecology; geology; limnology; environmental chemistry and physics; physical oceanography, meteorology and climatology, radiology, public health aspects of pollutant emissions, and systems engineering and integration.

WESD conducts broad environmental surveys, environmental program planning and data interpretation, and provides recommended action programs for meeting federal, state, and local environmental quality regulations. As a technical consultant on the Braidwood project, WESD staff biologists conducted a 2-year baseline study of the Braidwood Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Kankakee River in the vicinity of the site were determined, and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation in the biotic communities of the site were predicted.

B/B-UFSAR 1.4-5 1.4.5.9 Illinois Natural History Survey (INHS)

The Illinois Natural History Survey (INHS), which has its beginnings almost 120 years ago, is a division of the State Department of Registration and Education and provides services to farmers, homeowners, sportsmen, and all other citizens of Illinois as well as to industries. INHS cooperates in biological research with the Illinois Department of Agriculture, Conservation, and Public Health; the University of Illinois, Southern Illinois University, and other educational institutions; various research branches of the federal government; and other agricultural, conservation, municipal, and business organizations throughout the state.

INHS aquatic biologists were involved in a 4-year study of the Kankakee River and Horse Creek near Custer Park, Illinois. The purpose of the study is to obtain biological, physical, and chemical data which will be used to evaluate any effects of the construction and operation of the Braidwood Station and its associated cooling lake on the biota and water quality of the Kankakee River and Horse Creek. The station's cooling pond will use the Kankakee River as a source of water for both intake and discharge purposes.

1.4.5.10 NUS Corporation NUS Corporation is a consulting engineering, research, and testing firm specializing in environmental and energy systems engineering, systems analysis, design engineering, management consulting, and training programs related to these areas. NUS has provided advice and professional guidance to utility, industrial, and government clients throughout the United States and in a number of foreign countries.

As a technical consultant on the Braidwood project, NUS was involved in a study to determine the adequacy of the station's ultimate heat sink.

1.4.5.11 Eberline Instrument Corporation (EIC)

Eberline Instrument Corporation (EIC) has provided radiation measurement equipment, comprehensive radiation protection services, and analytical laboratory services to the nuclear industry since 1953.

As a technical consultant on the Byron/Braidwood projects EIC performed preoperational environmental radiological baseline studies on and around the site.

1.4.5.12 Meteorology Research, Inc. (MRI)

Meteorology Research, Inc. (MRI) is an environmental consulting firm which, since 1951, has provided meteorological and air

B/B-UFSAR 1.4-6 REVISION 1 - DECEMBER 1989 quality instruments and services to all aspects of industry in the solution of weather-related problems. These range from environmental impact assessments of existing or proposed airports and other major developments to problems of warehousing and marketing seasonal consumer goods. Of particular interest is the influence the local topography has on temperatures and winds.

MRI provided meteorological data from 1973 through mid-1975 for Byron and Braidwood Stations by implementation of an onsite meteorological measurement program.

1.4.5.13 Illinois State Museum (ISM)

The Illinois State Museum conducts archaeological investigations throughout the state of Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing and excavating, and laboratory analyses of datifacts, pollen, and soils.

As a technical consultant on the Braidwood project, ISM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Braidwood Station and pipeline corridor property.

1.4.5.14 Equitable Environmental Health, Inc. (EEH)

Equitable Environmental Health, Inc. (EEH), successor to Environmental Analysts, Inc./Tabershaw-Cooper Associated, Inc.,

is a multidisciplinary firm that offers the consulting services of medical professionals, scientists, engineers, economists, and technical support personnel in all areas of environmental health and economics.

EEH staff biologists conducted a 2-year baseline study of the Byron Station site. Distributions of phytoplankton, zooplankton, periphyton, benthos, fish, fish eggs and larvae, and water chemistry in the Rock River in the vicinity of the site were determined and quantitative data on terrestrial flora and fauna were collected. The impacts of plant construction and operation on the biotic communities of the site were predicted, and data were provided for a benefit-cost analysis of the project.

1.4.5.15 Espey, Huston & Associates, Inc. (EH & A)

Espey, Huston & Associates, Inc. (EH & A) is a consulting firm addressing the environmental problems associated with industrial and urban development. EH & A professionals cover a broad range of expertise including civil engineering, environmental engineering, mathematics and computer science, and all phases of aquatic, estuarine, and terrestrial ecology.

B/B-UFSAR 1.4-7 REVISION 1 - DECEMBER 1989 As a technical consultant on the Byron project, EH & A conducted the construction phase terrestrial and aquatic monitoring programs.

1.4.5.16 University of Wisconsin-Milwaukee (UWM)

The University of Wisconsin-Milwaukee under Dr. Elizabeth Benchley of the Dept. of Anthropology, conducts archaeological investigations throughout Wisconsin and northern Illinois. As a member of the Illinois Archaeological Survey, they have the expertise and services to perform contract archaeological work. Their studies included a pedestrian reconnaissance survey, subsurface testing, and lab analysis of datifacts, pollen, and soils.

As a technical consultant on the Byron project, UWM identified and made recommendations which Commonwealth Edison acted upon to aid in preserving the archaeological sites on Byron Station and pipeline corridor property. Also, UWM conducted archaeological investigations on the Byron transmission line right-of-ways.

1.4.5.17 Aero-Metric Engineering, Inc. (AME)

Aero-Metric Engineering, Inc., founded in 1969, is based in Sheboygan, Wisconsin. The staff was made up of over 50 technical photogrammetric personnel, many having professional engineer and/or survey registration. AME's capabilities allow for a complete range of precision photogrammetric services, including aerial photography, mapping, and multiple survey skills.

As a technical consultant on the Byron project, AME will be providing annual aerial infra-red photographs.

1.4.5.18 Iowa Institute of Hydraulic Research The Iowa Institute of Hydraulic Research, formally organized in 1931, is a Division of the University of Iowa's College of Engineering. The Institute staff exceeded 80 in number and was comprised of a professional staff with Ph.Ds in the areas of Civil Engineering, Mechanical Engineering, Physics, Mechanics and Hydraulics, and Aeronautical Engineering, with most of these personnel holding joint academic appointments in the College of Engineering's Division of Energy Engineering. The Institute of Hydraulic Research conducts programs of fundamental research and advanced design and analysis in the areas of environmental pollution, bioengineering, naval hydrodynamics, river mechanics, ice hydraulics, hydrology, water resources, hydraulic structures, fluid mechanics, advanced instrumentation and data-handling techniques for fluids research, and mathematical modeling of watersheds and hydrology.

B/B-UFSAR 1.4-8 REVISION 9 - DECEMBER 2002 As a technical consultant on the Braidwood project, the Institute conducted a thermal evaluation to determine the adequacy of the ultimate heat sink.

1.4.5.19 Babcock and Wilcox International (B&W)

B&W is located in Cambridge, Ontario, Canada. B&W has fabricated fossil-fueled boiler components for over 100 years and has fabricated nuclear system components since the late 1950's. B&W has supplied replacement steam generators for Byron Unit 1 and Braidwood Unit 1.

1.4.5.20 Framatome Technologies, Incorporated (FTI)

FTI is located in Lynchburg, Virginia and has been providing services to the electric power industry for over four decades.

FTI engineering services include the necessary expertise, experience, and NRC-approved computer codes and methodologies to support the transient analysis of the Unit steam generators.

1.4.5.21 Stone & Webster Engineers and Constructors, Inc. (S&W)

S&W is located in Boston, Massachusetts and has been providing services to the electric power industry for over 100 years. S&W has provided balance-of-plant design-engineering support services in support of the power uprate of the Byron and Braidwood units.

B/B-UFSAR 1.4-9 REVISION 8 - DECEMBER 2000 TABLE 1.4-1 EXELON GENERATION COMPANY'S NUCLEAR POWER PLANTS IN SERVICE OR UNDER CONSTRUCTION UNIT NOMINAL GROSS1 RATING (MWe)

SCHEDULED COMMERCIAL SERVICE DATE Dresden 1 210 1960 Dresden 2 850 1972 Dresden 3 850 1972 Quad-Cities 1 850 1972 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 La Salle 1 1122 1978 La Salle 2 1122 1979 Byron 1 1175 1985 Byron 2 1175 1987 Braidwood 1 1175 1988 Braidwood 2 1175 1988 1Note that this is a gross rating, not a net rating.

B/B-UFSAR 1.4-10 REVISION 1 - DECEMBER 1989 TABLE 1.4-2 NUCLEAR POWER PLANTS COMPLETED OR CURRENTLY UNDER DESIGN BY SARGENT & LUNDY UNIT NOMINAL GROSS2 RATING (MWe)

YEAR OF POWER OPERATION EBWR 5

1956 Elk River 22 1962 La Crosse 60 1967 SEFOR 20 (MWt) 1969 Dresden 2 850 1969 Dresden 3 850 1971 Quad-Cities 1 850 1971 Quad-Cities 2 850 1972 Zion 1 1085 1973 Zion 2 1085 1973 Fort St. Vrain, Unit 1 330 1973 Enrico Fermi, Unit 2 1200 1988 La Salle County Station, Unit 1 1122 1979 La Salle County Station, Unit 2 1122 1980 Byron Station, Unit 1 1175 1985 Byron Station, Unit 2 1175 1987 Braidwood Station, Unit 1 1175 1988 Braidwood Station, Unit 2 1175 1988 Clinton Power Station, Unit 1 992 1981 Kaiseraugst 992 1982 2Note that this is a gross rating, not a net rating.

B/B-UFSAR 1.4-11 TABLE 1.4-3 WESTINGHOUSE PRESSURIZED WATER REACTOR NUCLEAR POWER PLANTS PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Shippingport Duquesne Light Company; Energy Research & Development Administration Pennsylvania 1957 90 4

Yankee-Rowe Yankee Atomic Electric Company Massachusetts 1961 175 4

Trio Vercellese (Enrico Fermi)

Ente Nazionale per L'Energia Elettrica (ENEL)

Italy 1965 260 4

Chooz (Ardennes)

Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA)

France 1967 305 4

San Onofre Unit 1 Southern California Edison Co.;

San Diego Gas and Electric Co.

California 1968 450 3

Haddam Neck (Connecticut Yankee)

Connecticut Yankee Atomic Power Company Connecticut 1968 575 4

Jose Cabrera-Zorita Union Electrica, S.A.

Spain 1969 153 1

Beznau Unit 1 Nordostschweizerische Krafwerke AG (NOK)

Switzerland 1969 350 2

Robert Emmett Ginna Rochester Gas and Electric Corporation New York 1970 490 2

Mihama Unit 1 The Kansai Electric Power Company, Inc.

Japan 1970 320 2

Point Beach Unit 1 Wisconsin Electric Power Co.;

Wisconsin Michigan Power Co.

Wisconsin 1970 497 2

H. B. Robinson Unit 2 Carolina Power and Light Co.

South Carolina 1971 707 3

B/B-UFSAR 1.4-12 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beznau Unit 2 Nordostschweizerische Kraftwerke AG (NOK)

Switzerland 1972 350 2

Point Beach Unit 2 Wisconsin Electric Power Co.;

Wisconsin Michigan Power Co.

Wisconsin 1972 497 2

Surry Unit 1 Virginia Electric and Power Co.

Virginia 1972 822 3

Turkey Point Unit 3 Florida Power and Light Co.

Florida 1972 745 3

Indian Point Unit 2 Consolidated Edison Company of New York, Inc.

New York 1973 873 4

Prairie Island Unit 1 Northern States Power Company Minnesota 1973 530 2

Turkey Point Unit 4 Florida Power and Light Co.

Florida 1973 745 3

Surry Unit 2 Virginia Electric and Power Co.

Virginia 1973 822 3

Zion Unit 1 Exelon Generation Company Illinois 1973 1050 4

Kewaunee Wisconsin Public Service Corp.;

Wisconsin Power and Light Co.;

Madison Gas and Electric Co.

Wisconsin 1974 560 2

Prairie Island Unit 2 Northern States Power Company Minnesota 1974 530 2

Takahama Unit 1 The Kansai Electric Power Company, Inc.

Japan 1974 781 3

Zion Unit 2 Exelon Generation Company Illinois 1974 1050 4

B/B-UFSAR 1.4-13 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Doel Unit 1 Indivision Doel Belgium 1975 390 2

Doel Unit 2 Indivision Doel Belgium 1975 390 2

Donald C. Cook Unit 1 Indiana and Michigan Electric Company (AEP)

Michigan 1975 1060 4

Ringhals Unit 2 Statens Vattenfallsverk (SSPB)

Sweden 1975 822 3

Almaraz Unit 1 Unit Electrica, S.A.;

Compania Sevillana de Electricidad, S.A.;

Hidroelectrica Espanola, S.A.

Spain 1976 902 3

Beaver Valley Unit 1 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Company Pennsylvania 1976 852 3

Diablo Canyon Unit 1 Pacific Gas and Electric Co.

California 1976 1084 4

Indian Point Unit 3 Consolidated Edison Company of New York, Inc.

New York 1976 965 4

Lemoniz Unit 1 Iberduero, S.A.

Spain 1976 902 3

Salem Unit 1 Public Service Electric and Gas Company; Exelon Generation Company; Atlantic City Electric Co.;

Delmarva Power and Light Co.

New Jersey 1976 1090 4

B/B-UFSAR 1.4-14 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Trojan Portland General Electric Co.;

Eugene Water and Electric Board; Pacific Power and Light Company Oregon 1976 1130 4

Almaraz Unit 2 Union Electrica, S.A.;

Compania Sevillana de Electricidad, S.A.;

Hidroelectrica Espanola, S.A.

Spain 1977 902 3

Asco Unit 1 Fuerzas Electricas de Cataluna, S.A. (FESCA)

Spain 1977 902 3

Diablo Canyon Unit 2 Pacific Gas and Electric Co.

California 1977 1106 4

Joseph M. Farley Unit 1 Alabama Power Company Alabama 1977 829 3

Ko-Ri Unit 1 Korea Electric Power Co., Ltd.

Korea 1977 564 2

North Anna Unit 1 Virginia Electric and Power Co.

Virginia 1977 898 3

North Anna Unit 2 Virginia Electric and Power Co.

Virginia 1977 898 3

Ohi Unit 1 The Kansai Electric Power Co., Inc.

Japan 1977 1122 4

Ohi Unit 2 The Kansai Electric Power Co., Inc.

Japan 1977 1122 4

Ringhals Unit 3 Statens Vattenfallsvert (SSPB)

Sweden 1977 900 3

Sequoyah Unit 1 Tennessee Valley Authority Tennessee 1977 1148 4

Angra dos Reis Unit 1 Furnas-Centrais Electricas, S.A.

Brazil 1978 626 2

B/B-UFSAR 1.4-15 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Asco Unit 2 Fuerzas Electricas de Cataluna, S.A. (FESCA);

Empresa Nacional Hidroelectrica del Ribagorzana, S.A. (ENHER);

Fuerzas Hidroelectricas del Segre, S.A.;

Hidroelectrica de Cataluna, S.A.

Spain 1978 902 3

Donald C. Cook Unit 2 Indiana and Michigan Electric Company (AEP)

Michigan 1978 1060 4

Lemoniz Unit 2 Iberduero, S.A.

Spain 1978 902 3

Sequoyah Unit 2 Tennessee Valley Authority Tennessee 1978 1148 4

Watts Bar Unit 1 Tennessee Valley Authority Tennessee 1978 1177 4

William B. McGuire Unit 1 Duke Power Company North Carolina 1978 1180 4

Joseph M. Farley Unit 2 Alabama Power Company Alabama 1979 829 3

Krsko Savske Elektrarne, Ljubljana, Slovenia, Elektroprivreda, Zagreb, Croatia Yugoslavia 1979 615 2

Ringhals Unit 4 Statens Vattenfallsvert (SSPD)

Sweden 1979 900 3

B/B-UFSAR 1.4-16 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Salem Unit 2 Public Service Electric and Gas Company; Exelon Generation Company Atlantic City Electric Co.;

Delmarva Power and Light Co.

New Jersey 1979 1115 4

Virgil C. Summer South Carolina Electric and Gas Company South Carolina 1979 900 3

Watts Bar Unit 2 Tennessee Valley Authority Tennessee 1979 1177 4

William B. McGuire Unit 2 Duke Power Company North Carolina 1979 1180 4

Byron Unit 1 Exelon Generation Company Illinois 1981 1120 4

Comanche Peak Unit 1 Texas Utilities Generating Co.

Texas 1980 1150 4

Seabrook Unit 1 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1980 1200 4

South Texas Project Unit 1 Houston Lighting and Power Co.;

Central Power and Light Co.;

City Public Service of San Antonio; City of Austin, Texas Texas 1980 1250 4

B/B-UFSAR 1.4-17 REVISION 8 - DECEMBER 2000 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Beaver Valley Unit 2 Duquesne Light Company; Ohio Edison Company; Pennsylvania Power Co.;

Cleveland Electric Illuminating Company; Toledo Edison Company Pennsylvania 1981 852 3

Braidwood Unit 1 Exelon Generation Company Illinois 1981 1120 4

Callaway Unit 1 SNUPPS - Union Electric Co.

Missouri 1981 1150 4

Catawba Unit 1 Duke Power Company South Carolina 1981 1153 4

Jamesport Unit 1 Long Island Lighting Company New York 1981 1150 4

Ko-Ri Unit 2 Korea Electric Power Co., Ltd.

Korea 1981 605 2

NORCO Puerto Rico Water Resources Authority Puerto Rico 583 2

Braidwood Unit 2 Exelon Generation Company Illinois 1982 1120 4

Byron Unit 2 Exelon Generation Company Illinois 1982 1120 4

Catawba Unit 2 Duke Power Company South Carolina 1982 1153 4

Comanche Peak Unit 2 Texas Utilities Generating Co.

Texas 1982 1150 4

Marble Hill Unit 1 Public Service Company of Indiana, Inc.;

Northern Indiana Public Service Company Indiana 1982 1150 4

B/B-UFSAR 1.4-18 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Millstone Unit 3 Northeast Nuclear Energy Co.

Connecticut 1982 1156 4

Seabrook Unit 3 Public Service Company of New Hampshire; United Illuminating Company New Hampshire 1982 1200 4

South Texas Project Unit 2 Houston Lighting and Power Co.;

Central Power and Light Co.;

City Public Service of San Antonio; City of Austin, Texas Texas 1982 1250 4

Taiwan Unit 5 Taiwan Power Company Taiwan 1982 950 3

Wolf Creek Unit 1 SNUPPS - Kansas Gas and Electric Company; Kansas City Power and Light Company Kansas 1982 1150 4

Alvin W. Vogtle Unit 1 Georgia Power Company Georgia 1983 1113 4

Callaway Unit 2 SNUPPS - Union Electric Company Missouri 1983 1150 4

NEP-1 New England Power Company 1983 1150 4

Fort Calhoun Unit 2 Omaha Public Power District; Nebraska Public Power District Nebraska 1983 1150 4

Jamesport Unit 2 Long Island Lighting Company New York 1983 1150 4

Sears Island Central Maine Power Company Maine 1200 4

B/B-UFSAR 1.4-19 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Taiwan Unit 6 Taiwan Power Company Taiwan 1983 950 3

Alvin W. Vogtle Unit 2 Georgia Power Company Georgia 1984 1113 4

Marble Hill Unit 2 Public Service Company of Indiana, Inc.;

Northern Indiana Public Service Company Indiana 1984 1150 4

Shearon Harris Unit 1 Carolina Power and Light Co.

North Carolina 1984 900 3

Sterling SNUPPS - Rochester Gas and Electric Corporation; Central Hudson Gas and Electric Corporation; Niagara Mohawk Power Corporation; Orange and Rockland Utilities, Inc.

New York 1984 1150 4

Atlantic Unit 1 (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Co.;

Jersey Central Power and Light Company New Jersey 1985 1150 4

NEP-2 New England Power Company 1985 1150 4

South Dade Unit 1 Florida Power and Light Co.

Florida 1985 1150 4

Sundesert Unit 1 San Diego Gas and Electric Co.

California 1985 950 3

B/B-UFSAR 1.4-20 TABLE 1.4-3 (Cont'd)

PLANT OWNER UTILITY LOCATION SCHEDULED COMMERCIAL OPERATION MWe NET NUMBER OF LOOPS Tyrone Unit 1 SNUPPS - Northern States Power Company Wisconsin 1985 1150 4

Shearon Harris Unit 2 Carolina Power and Light Co.

North Carolina 1986 900 3

South Dade Unit 2 Florida Power and Light Co.

Florida 1986 1150 4

Atlantic No. (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Co.;

Jersey Central Power and Light Company New Jersey 1987 1150 4

Shearon Harris Unit 4 Carolina Power and Light Co.

North Carolina 1988 900 3

Sundesert Unit 2 San Diego Gas and Electric Co.

California 1988 950 3

Sayago Unit 1 Iberduero, S.A.

Spain 1980's 1000 3

Sayago Unit 4 Iberduero, S.A.

Spain 1980's 1000 3

Shearon Harris Unit 3 Carolina Power and Light Co.

North Carolina 1990 900 3

Unassigned Unit 1 (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1990 1150 4

Unassigned Unit 2 (O.P.S.)

Public Service Electric and Gas Company; Atlantic City Electric Company New Jersey 1992 1150 4

B/B-UFSAR 1.5-1 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION The design of the Byron/Braidwood units is based upon proven concepts which have been developed and successfully applied to the design of pressurized water reactor systems. There are currently no areas of research and development which are required for operation of this plant.

At the time of issuance of construction permits for the Byron/

Braidwood units, the Preliminary Safety Analysis Report (PSAR) and the standard design report which it referenced, RESAR-3, identified certain research and development programs which were incomplete. These programs, which have been successfully completed, have provided technical information which has been used either to demonstrate the safety of design, more sharply define margins of conservatism, or lead to design improvements.

Reference 1 presents descriptions of those safety-related research and development programs which have been carried out for, by, or in conjunction with Westinghouse Nuclear Energy Systems, and which are applicable to Westinghouse pressurized water reactors. The discussion which follows documents the completion of the construction permit stage research programs.

1.5.1 Programs Required for Plant Operation Two programs were identified as required for plant design and operation in the PSAR:

a. core stability evaluation and
b. fuel rod burst program.

Both programs are complete. The fuel rod burst program was completed at the time of the PSAR. The core stability evaluation program was not. A discussion of the core stability evaluation program follows.

1.5.1.1 Core Stability Evaluation The program to establish means for the detection and control of potential xenon oscillations and for the shaping of the axial power distribution for improved core performance has been satisfactorily completed. See item 1, Reference 2, for a further discussion of the tests and results.

1.5.2 Other Programs Not Required for Plant Operation The following programs were not complete at the time of the PSAR but are now satisfactorily complete.

B/B-UFSAR 1.5-2 1.5.2.1 Fuel Development Program for Operation at High Power Densities The program to demonstrate the satisfactory operation of fuel at high burnup and power densities has been satisfactorily completed. See item 8, Reference 2, for a further discussion of the program and its results.

1.5.2.2 Blowdown Forces Program Westinghouse has completed BLODWN-2, an improved digital computer program for the calculation of local fluid pressures, flows and density transients in the primary coolant systems during a LOCA.

BLODWN-2 is used to evaluate the effects of blowdown forces in this application. Refer to item 15 in Reference 4 for a further discussion of the tests and results.

1.5.2.3 Blowdown Heat Transfer Testing (Formerly Titled Delayed Departure From Nucleate Boiling)

The NRC Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Power Reactors was issued in Section 50.46 of 10 CFR 50 on December 28, 1973. It defines the basis and conservative assumptions to be used in the evaluation of the performance of emergency core cooling systems (ECCS).

Westinghouse believes that some of the conservatism of the criteria is associated with the manner in which transient DNB phenomena are treated in the evaluation models. Transient critical heat flux data presented at the 1972 specialists meeting of the Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be delayed under transient conditions.

To demonstrate the conservatism of the ECCS evaluation models, Westinghouse initiated a program to experimentally simulate the blowdown phase of a LOCA. This testing is part of the Electric Power Research Institute (EPRI) sponsored Blowdown Heat Transfer Program, which was started early in 1976. Testing was completed in 1979. A DNB correlation developed by Westinghouse from these test results is used in the ECCS analyses for Byron/Braidwood.

Objective The objective of the blowdown heat transfer test was to determine the time that DNB occurs under LOCA conditions. This information was used to confirm a new Westinghouse transient DNB correlation.

The steady-state DNB data obtained from 15x15 and 17x17 test programs was used to assure that the geometrical differences between the two fuel arrays is correctly treated in the transient correlations.

B/B-UFSAR 1.5-3 Program The program was divided into two phases. The Phase I tests started from steady-state conditions, with sufficient power to maintain nucleate boiling throughout the bundle, and progressed through controlled ramps of decreasing test section pressure or flow initiated DNB. By applying a series of controlled conditions, investigation of the DNB was studied over a range of qualities and flows, and at pressures relevant to a PWR blowdown.

Phase I provided separate-effects data for heat transfer correlation development.

Typical parameters used for Phase I testing are shown in Table 1.5-1.

Phase II simulated PWR behavior during a LOCA to permit definition of the time delay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after establishment of typical steady-state operating conditions. The fluid transient was then initiated, and the rod power decay was programmed in such a manner as to simulate the actual heat input of fuel rods. The test was terminated when the heater rod temperatures reached a predetermined limit.

Typical parameters used for Phase II testing are shown in Table 1.5-2.

Test Description The experimental program was conducted in the J-Loop at the Westinghouse Forest Hills Facility with a full length 5x5 rod bundle simulating a section of a 15x15 fuel assembly to determine DNB occurrence under LOCA conditions.

The heater rod bundles used in this program were internally-heated rods, capable of a maximum linear power of 18.8 kW/ft, with a total power of 135 kW (for extended periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped cosine power distribution. Each rod was adequately instrumented with a total of 12 clad thermocouples.

Results The experiments in the DNB facility resulted in cladding temperature and fluid properties measured as a function of time throughout the blowdown range from 0 to 20 seconds.

Facility modifications and installation of the initial test bundle were completed. A series of shakedown tests in the

B/B-UFSAR 1.5-4 REVISION 1 - DECEMBER 1989 J-Loop were performed. These tests provided data for instrumentation calibration and check-out, and provided information regarding facility control and performance. Initial program tests were performed during the first half of 1975.

Under the sponsorship of EPRI, testing was reinitiated during 1976 on the same test bundle. The testing was terminated in November 1976 and plans were made for a new test bundle and further testing during 1978-1979. These tests were completed in December of 1979.

1.5.3 References

1.

F. T. Eggleston, "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, October 1978.

2.

F. T. Eggleston, "Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8768.

Spring 1976 Edition.

3.

"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8458. Fall 1977 Edition.

4.

"Safety-Related Research and Development for Westinghouse PWRs Program Summaries," WCAP-8004. Fall 1972 Edition.

B/B-UFSAR 1.5-5 TABLE 1.5-1 BLOWDOWN HEAT TRANSFER PHASE I TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 1250 to 2250 psia Test section mass velocity 1.12 to 2.5x106 lb/hr-ft2 Core inlet temperature 550° F to 600° F Maximum heat flux 306,000 to 531,000 Btu/hr-ft2 TRANSIENT RAMP CONDITIONS Pressure decrease 0 to 350 psia/sec and subcooled depressurization from 2250 psia Flow decrease 0 to 100%/sec Inlet enthalpy constant

B/B-UFSAR 1.5-6 TABLE 1.5-2 BLOWDOWN HEAT TRANSFER PHASE II TEST PARAMETERS PARAMETERS NOMINAL VALUE INITIAL STEADY-STATE CONDITIONS Pressure 2250 psia Test section mass velocity 2.5x106 lb/hr-ft2 Inlet coolant temperature 545° F Maximum heat flux 531,000 Btu/hr-ft2 TRANSIENT CONDITIONS Simulated break Double-ended cold leg guillotine breaks

B/B-UFSAR 1.6-1 1.6 MATERIAL INCORPORATED BY REFERENCES Table 1.6-1 lists topical reports which provide information additional to that provided in this UFSAR and which have been filed separately with the Nuclear Regulatory Commission (NRC) in support of this and similar applications.

A legend to the review status code letters follows:

A

- NRC review complete; NRC acceptance letter issued.

AE

- NRC accepted as part of the Westinghouse Emergency Core Cooling System (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.

B

- Submitted to the NRC as background information; not undergoing formal NRC review.

O

- On file with NRC; older generation report with current validity; not actively under formal NRC review.

U

- Actively under formal NRC review.

B/B-UFSAR 1.6-2 TABLE 1.6-1 TOPICAL REPORTS INCORPORATED BY REFERENCE REPORT REFERENCE SECTION(S)

REVIEW STATUS "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," WCAP-2048, July 1962 4.3 0

"Single Phase Local Boiling and Bulk Boiling Pressure Drop Correlations," WCAP-2850 (Proprietary), April 1966 and WCAP-7916 (Non-Proprietary), June 1972 4.4 0

"In-Pile Measurement of UO2 Thermal Conductivity," WCAP-2923, 1966 4.4 0

"Hydraulic Tests of the San Onofre Reactor Model," WCAP-3269-8, June 1964 4.4 0

"LEOPARD - A Spectrum Dependent Non-Spatial Depletion Code for the IBM - 7094,"

WCAP-3269-26, September 1963 4.3, 4.4 15.0, 15.4 0

"Saxton Core II Fuel Performance Evaluation,"

WCAP-3385-56, Part II, "Evaluation of Mass Spectrometric and Radiochemical Analyses of Irradiated Saxton Plutonium Fuel," July 1970 4.3, 4.4 0

"Xenon-Induced Spatial Instabilities in Large PWRs," WCAP-3680-20, (EURAEC-1974)

March 1968 4.3 0

"Control Procedures for Xenon-Induced X-Y Instabilities in Large PWR's,"

WCAP-3680-21, (EURAEC-2111) February 1969 4.3 0

"Xenon-Induced Spatial Instabilities in Three-Dimensions," WCAP-3680-22, (EURAEC-2116) September 1969 4.3 0

"Pressurized Water Reactor pH - Reactivity Effect Final Report," WCAP-3698-8, (EURAEC-2074) October 1968 4.3 0

"PUO2 - UO2 Fueled Critical Experiments,"

WCAP-3726-I, July 1967 4.3 0

B/B-UFSAR 1.6-3 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Melting Point of Irradiated UO2,"

WCAP-6065, February 1965 4.2, 4.4 0

"Burnup Physics of Heterogeneous Reactor Lattices," WCAP-6069, June 1965 4.4 0

"LASER - A Depletion Program for Lattice Calculations Based on MUFT and THERMOS,"

WCAP-6073, April 1966 4.3 0

"Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium," WCAP-6086, August 1969 4.3 0

"Subchannel Thermal Analysis of Rod Bundle Cores," WCAP-7015, Revision 1, January 1969 4.4 0

"The PANDA Code," WCAP-7048 (Proprietary) and WCAP-7757 (Non-Proprietary), January 1975 4.3 A

"Evaluation of Protective Coatings for Use in Reactor Containment," WCAP-7198-L (Proprietary), April 1969 and WCAP-7825 (Non-Proprietary), December 1971 4.3 0

"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7208 (Proprietary), September 1968 and WCAP-7811, (Non-Proprietary), December 1971 4.3 "The TURTLE 24.0 Diffusion Depletion Code,"

WCAP-7213 (Proprietary) and WCAP-7758 (Non-Proprietary), January 1975 4.3, 15.0 15.4 A

"Core Power Capability in Westinghouse PWRs,"

WCAP-7267-L (Proprietary), October 1969 and WCAP-7809 (Non-Proprietary), December 1971 4.3 "Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors,"

WCAP-7306, April 1969 15.4 "Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7308, December 1971 4.3 A

B/B-UFSAR 1.6-4 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Application of the THINC Program to PWR Design," WCAP-7359-L (Proprietary), August 1969 and WCAP-7838 (Non-Proprietary),

January 1972 4.4 O

"Seismic Testing of Electrical and Control Equipment," WCAP-7397-L (Proprietary) and WCAP-7817 (Non-Proprietary), December 1971 3.10 O

"Seismic Testing of Electrical and Control Equipment (WCID Process Control Equipment),"

WCAP-7397-L, Supplement 1 (Proprietary) and WCAP-7817, Supplement 1 (Non-Proprietary),

December 1971 3.10 O

"Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," WCAP-7477-L (Proprietary), March 1970 and WCAP-7735 (Non-Proprietary), August 1971 5.2 A

"Radiological Consequences of a Fuel Handling Accident," WCAP-7518-L (Proprietary) and WCAP-7828 (Non-Proprietary), June 1970 15.7 O

"Seismic Vibration Testing with Sine Beats,"

WCAP-7558, October 1972 3.10 O

"An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975 15.4 A

"Dynamic Fracture Toughness Properties of Heavy Section A533 Grade B Class 1 Steel Plate," WCAP-7623, December 1970 5.4 O

"Interchannel Thermal Mixing with Mixing Vane Grids," WCAP-7667-L (Proprietary) and WCAP-7755 (Non-Proprietary), January 1975 4.4 A

"DNB Tests Results for New Mixing Vane Grids (R)," WCAP-7695-L (Proprietary) and WCAP-7958 (Non-Proprietary) and Addendum, January 1975 4.4 A

B/B-UFSAR 1.6-5 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "An Evaluation of Solid State Logic Reactor Protection in Anticipated Transients,"

WCAP-7706, February 1973 4.6, 7.1 O

"Electric Hydrogen Recombiner for PWR Containments," WCAP-7709-L, Supplements 1 through 7 (Proprietary) and WCAP-7820, Supplements 1 through 7 (Non-Proprietary),

1971 through 1977 3.11, 6.2 A

"A Comprehensive Space-Time Dependent Analysis of Loss of Coolant (SATAN-IV Digital Code),"

WCAP-7750, August 1971 3.6 O

"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, October 1971 15.2 O

"Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Revision 1, June 1972 5.2 O

"Behavior of Austenitic Stainless Steel in Post Hypothetical Loss of Coolant Accident Environment," WCAP-7798-L (Proprietary) and WCAP-7803 (Non-Proprietary), January 1972 6.1 O

"Nuclear Fuel Division Quality Assurance Program Plan," WCAP-7800, Revision 4-A, April 1975 4.2, 17 A

"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"

WCAP-7806, December 1971 4.3 B

"Power Distribution Control of Westinghouse Pressurized Water Reactors," WCAP-7811, December 1971 4.3 O

"Seismic Testing of Electrical and Control Equipment (Low Seismic Plants)," WCAP-7817, Supplements 1-8, December 1971-March 1974 3.10 O

B/B-UFSAR 1.6-6 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Evaluation of Steam Generator Tube, Tubesheet and Divider Plate Under Combined LOCA Plus SSE Conditions," WCAP-7832, December 1973 5.4 A

"Inlet Orificing of Open PWR Cores,"

WCAP-7836, January 1972 4.4 B

"Neutron Shielding Pads," WCAP-7870, May 1972 3.9 A

"LOFTRAN Code Description," WCAP-7907, June 1972 5.2, 15.0 15.1, 15.2, 15.3, 15.4, 15.5, 15.6 A

"FACTRAN - A FORTRAN-IV Code for Thermal Transients in a UO2 Fuel Rod," WCAP-7908, June 1972 15.0, 15.2 15.3, 15.4 A

"MARVEL, A Digital Computer Code for Transient Analysis of a Multiloop PWR System,"

WCAP-7909, June 1972 6.3 O

"Power Peaking Factors," WCAP-7912-L (Proprietary) and WCAP-7912 (Non-Proprietary), January 1975 and Supplement 4.3, 4.4 A

"Damping Values of Nuclear Power Plant Components," WCAP-7921, May 1974 lA, 3.7 A

"Basis for Heatup and Cooldown Limit Curves," WCAP-7924, April 1975 5.3 A

"Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid,"

WCAP-7941-L (Proprietary) and WCAP-7959 (Non-Proprietary), January 1975 4.4 A

"Fuel Assembly Safety Analysis for Combined Seismic and Loss of Coolant Accident, 15x15,"

WCAP-7950, July 1972 3.7 A

"THINC-IV An Improved Program for Thermal and Hydraulic Analysis of Rod Bundle Cores,"

WCAP-7956, June 1973 4.4 A

B/B-UFSAR 1.6-7 REVISION 9 - DECEMBER 2002 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Axial Xenon Transient Tests at the Rochester Gas and Electric Reactor," WCAP-7964, June 1971 4.3 O

"TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979 (Proprietary) and WCAP-8028 (Non-Proprietary), January 1975 15.0, 15.4 A

"WIT-6 Reactor Transient Analysis Computer Program Description," WCAP-7980, November 1972 15.0, 15.4 A

"Application of Modified Spacer Factor to "L" Grid Typical and Cold Wall Cell DNB,"

WCAP-7988 (Proprietary) and WCAP-8030 (Non-Proprietary), October 1972 4.4 A

"Application of the THINC-IV Program to PWR Design," WCAP-8054 (Proprietary) and WCAP-8195 (Non-Proprietary), October 1973 4.4 A

"Pipe Breaks for the LOCA Analysis of the Westinghouse Primary Coolant Loop,"

WCAP-8082 (Proprietary) and WCAP-8172 (Non-Proprietary), January 1975 3.6 A

"Reactor Coolant Pump Integrity in LOCA,"

WCAP-8163, September 1973 lA, 5.4 O

"Calculational Model for Core Reflooding After a Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974 15.6 A

"Effect of Local Heat Flux Spikes on DNB in Non-Uniform Heated Rod Bundles," WCAP-8174 (Proprietary) and WCAP-8202, (Non-Proprietary), August 1973 4.4 A

"WFLASH, A FORTRAN-IV Computer Program for Simulation of Transients in a Multi-Loop PWR," WCAP-8200, Revision 2 (Proprietary) and WCAP-8261, Revision 1 (Non-Proprietary), July 1974 15.6 A

B/B-UFSAR 1.6-8 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Fuel Densification Experimental Results and Model for Reactor Application,"

WCAP-8218 (Proprietary) and WCAP-8219 (Non-Proprietary), March 1975 4.1, 4.2, 4.3, 4.4 A

"Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236 (Proprietary), December 1973 and WCAP-8288 (Non-Proprietary), January 1974 and Addenda 3.7, 4.2 A

"Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," WCAP-8236, Addendum 1 (Proprietary), March 1974 and WCAP-8288, Addendum 1 (Non-Proprietary), April 1974 3.7 A

"Documentation of Selected Westinghouse Structural Analysis Computer Codes,"

WCAP-8252, Revision 1, July 1977 3.6, 3.9 O

"Hydraulic Flow Test of the 17x17 Fuel Assembly," WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary), February 1974 4.2, 4.4 O

"Effect of 17x17 Fuel Assembly Geometry on DNB," WCAP-8296 (Westinghouse Proprietary) and WCAP-8927 (Non-Proprietary), February 1975 4.4 A

"The Effect of 17x17 Fuel Assembly Geometry on Interchannel Thermal Mixing," WCAP-8298 (Proprietary) and WCAP-8299 (Non-Proprietary),

January 1975 4.4 A

"LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974 15.0, 15.6 AE SATAN-IV Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant,"

WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974 15.0, 15.6 AE

B/B-UFSAR 1.6-9 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests,"

WCAP-8303 (Proprietary) and WCAP-8317 (Non-Proprietary), July 1975 3.9 A

"Control of Delta Ferrite in Austenitic Stainless Steel Weldments," WCAP-8324-A, June 1974 lA, 5.2 A

"Containment Pressure Analysis Code (COCO),"

WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974 15.6 AE "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974 4.3, 4.6, 15.1, 15.2, 15.4, 15.8 O

"Westinghouse ECCS Evaluation Model -

Summary," WCAP-8339, July 1974 6.2, 15.6 AE "Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary) and WCAP-8356 (Non-Proprietary), July 1974 15.6 AE "Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP -8341 (Proprietary) and WCAP-8342 (Non-Proprietary), July 1974 lA(N), 17 A

"Effects of Fuel Densification Power Spikes on Clad Thermal Transients," WCAP-8359, July 1974 4.3 AE "Westinghouse Nuclear Energy Systems Division Quality Assurance Plan," WCAP-8370, Revision 9A, September 1977 1A, 17 A

"Qualification of Westinghouse Seismic Testing Procedure for Electrical Equipment Tested Prior to May 1974," WCAP-8373, August 1974 3.10 O

"Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-Proprietary),

July 1974 4.2 A

B/B-UFSAR 1.6-10 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Power Distribution Control and Load Following Procedures," WCAP-8385 (Proprietary) and WCAP-8403 (Non-Proprietary), September 1974 4.3, 4.4 A

"An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," WCAP-8424, Revision 1, June 1975 15.3 O

"17x17 Drive Line Components Tests - Phase IB, II, III, D-Loop Drop and Deflection,"

WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December 1974 3.9, 15.0 A

"Analysis of Data from the Zion (Unit 1)

THINC Verification Test," WCAP-8453-A (Proprietary), May 1976 and WCAP-8454 (Non-Proprietary), January 1975 4.4 A

"Westinghouse ECCS Evaluation Model -

Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Non-Proprietary), April 1974 15.6 AE "Incore Power Distribution Determination in Westinghouse Pressurized Water Reactors,"

WCAP-8498, July 1975 4.3 O

"UHI Plant Internals Vibration Measurement Program and Pre and Post Hot Functional Examinations," WCAP-8516-P (Proprietary) and WCAP-8517 (Non-Proprietary), April 1975 3.9 A

"Critical Heat Flux Testing of 17x17 Fuel Assembly Geometry with 22 Inch Spacing,"

WCAP-8536 (Proprietary) and WCAP-8537 (Non-Proprietary), May 1975 4.4 A

"Westinghouse ECCS - Four Loop Plant (17x17)

Sensitivity Studies," WCAP-8565 (Proprietary) and WCAP-8566 (Non-Proprietary), July 1975 15.6 A

B/B-UFSAR 1.6-11 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Improved Thermal Design Procedure,"

WCAP-8567-P, July 1975 (Proprietary) and WCAP-8568, July 1975 (Non-Proprietary) 4.4, 15.0 A

"Augmented Startup and Cycle 1 Physics Program Supplement 1," WCAP-8575, June 1976 (Proprietary) and WCAP-8576, June 1976 (Non-Proprietary) and Supplements.

4.3 O

"The Application of Preheat Temperatures After Welding Pressure Vessel Steels,"

WCAP-8577, February 1976 lA A

"Failure Mode and Effects Analysis (FMEA) of the Engineered Safeguard Features Actuation System," WCAP-8584 (Proprietary) and WCAP-8760 (Non-Proprietary), April 1976 4.6 O

"Environmental Qualification of Westinghouse NSSS Class lE Equipment," WCAP-8587, September 1975 lA, 3.1O, 3.11 A

"Westinghouse ECCS Evaluation Model - October 1975 Version," WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary), November 1975 15.6 A

"Experimental Verification of Wet Fuel Storage Criticality Analyses," WCAP-8682 (Proprietary) and WCAP-8683 (Non-Proprietary), December 1975 4.3 B

"Fuel Rod Bowing," WCAP-8691 (Proprietary) and WCAP-8692 (Non-Proprietary),

December 1975 4.2 O

"Delta Ferrite in Production Austenitic Stainless Steel Weldments," WCAP-8693, January 1976 lA, 5.2 B

"MULTIFLEX - A FORTRAN-IV Computer Program for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708 (Proprietary) and WCAP-8709 (Non-Proprietary), February 1976 3.9 A

B/B-UFSAR 1.6-12 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS Foster, J. P., et al., Westinghouse Improved Performance Analysis and Design Model (PAD 4.0), WCAP-15063-P-A, Revision 1 with Errata, July 2000.

4.2 A

"New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, July 1976 (Proprietary) and WCAP-8763, July 1976 (Non-Proprietary) 4.4 A

"Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries," WCAP-8768, Revision 2, October 1978 1.5, 4.2, 4.3 B

"Verification of Neutron Pad and 17x17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant," WCAP-8780, May 1976 3.9 B

"Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations,"

WCAP-8785, October 1976 4.2 "Hybrid B4C Absorber Control Rod Evaluation Report," WCAP-8846, October 1977 4.2, 15.0 15.3 A

"Westinghouse ECCS - Four Loop Plant (17x17)

Sensitivity Studies with Upper Head Fluid Temperature at Thot," WCAP-8865, May 1977 15.6 A

"7300 Series Process Control System Noise Tests," WCAP-8892-A, April 1977 7.1 A

"Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8963 (Proprietary), November 1976 and WCAP-8964 (Non-Proprietary), August 1977 4.2 A

"Westinghouse Emergency Core Cooling System Small Break October 1975 Model," WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary),

April 1977 15.6 A

B/B-UFSAR 1.6-13 REVISION 1 - DECEMBER 1989 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Failure Mode and Effects Analysis of the Solid State Full Length Rod Control System,"

WCAP-8976, September 1977 4.6 O

"Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods,"

WCAP-9000-L, Revision 1 (Proprietary), July 1969 and WCAP-7806 (Non-Proprietary), December 1971.

4.3 "Axial Power Distribution Monitoring Using Four-Section Ex-Core Detectors," WCAP-9105 (Proprietary) and WCAP-9106 (Non-Proprietary),

July 1977 4.3 A

"Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCAs During Operation with One Loop Out of Service for Plants Without Loop Isolation Valves,"

WCAP-9166 (Proprietary) and WCAP-9167 (Non-Proprietary), February 1978 15.6 O

"Westinghouse Emergency Core Cooling System Evaluation Model - Modified October 1975 Version," WCAP-9168 (Proprietary) and WCAP-9150 (Non-Proprietary), September 1977 15.6 O

"Properties of Fuel and Core Component Materials," WCAP-9179 (Proprietary), September 1977 and WCAP-9224 (Non-Proprietary) 4.2 O

"Westinghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220 (Proprietary Version), WCAP-9221 (Non-Proprietary Version), February 1978 15.6 A

"Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly," WCAP-9401 (Proprietary) and WCAP-9402 (Non-Proprietary), March 1979 4.1, 4.2, 4.4 A

"PALADON - Westinghouse Nodal Computer Code,"

WCAP-9485 (Proprietary) and WCAP-9486 (Non-Proprietary) December 1978 4.3 A

B/B-UFSAR 1.6-14 REVISION 15 - DECEMBER 2014 TABLE 1.6-1 (Cont'd)

REPORT REFERENCE SECTION(S)

REVIEW STATUS "Reference Core Report 17x17 Optimized Fuel Assembly," WCAP-9500 (Non Proprietary),

July 1979 4, 15 A

"RELAP5/MOD2-B&W - An Advanced Computer Code for Light Water Reactor LOCA and non-LOCA Transient Analysis" BAW-10164, Revision 3 (non-proprietary), October 1996 15 A

"CONTEMPT - Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident,", BAW-10095A, Revision 1, April 1978 6

O Beacon Core Monitoring and Operations Support System, WCAP-12472 (Proprietary Class 2),

August 1994 4.3, 4.4, 7.7 A

Relaxation of Constant Axial Offset Control, FQ Surveillance Technical Specification, WCAP-10216-P-A, Revision 1A (Proprietary Class 2), February 1994 4.3, 4.4 A

VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, WCAP-14565-P-A (Proprietary) / WCAP-15306-NP-A (Non-Proprietary), October 1999 4.4, 15.0 A

Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications, WCAP-14565-P-A Addendum 2-P-A (Proprietary) / WCAP-15306-NP-A Addendum 2-NP-A (Non-Proprietary), April 2008 4.4 A

SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, WCAP-10698-P-A (Proprietary) and WCAP-10750-A (Non-Proprietary), August 1987 15.0 A

B/B-UFSAR 1.7-1 REVISION 9 - DECEMBER 2002 1.7 DRAWINGS The drawings cited in each UFSAR Chapter 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. References on the figures contained in the UFSAR to ComEd, CECo, and Commonwealth Edison will be revised to reflect the change in facility ownership to Exelon Generation Company when other changes to that figure are needed.

1.7.1 Electrical, Instrumentation, and Control Drawings Subsection 1.7.1 of the FSAR provides a list of electrical, instrumentation, and control drawings that were provided to the NRC during the initial licensing phase.

1.7.2 Drawings for Independent Structural Review Subsection 1.7.2 of the FSAR provides a list of the structural, architectural, mechanical loading and electrical loading drawings that were provided to the NRC to enable them to perform the Project Structural Review and the Independent Structural Review during the licensing phase.

B/B-UFSAR 1.7-2 REVISION 9 - DECEMBER 2002 TABLE 1.7-1 This Table has been intentionally deleted.

B/B-UFSAR REVISION 9 - DECEMBER 2002 Pages 1.7-3 through 1.7-17 have been intentionally deleted.

B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.1-1 through 1.1-3 have been deleted intentionally.

B/B-UFSAR REVISION 9 - DECEMBER 2002 Figures 1.2-1 through 1.2-17 have been deleted intentionally.

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