RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 1 - Introduction and General Description of Plant

Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 1 - Introduction and General Description of Plant

ML14015A104
Person / Time
Site:	Clinton
Issue date:	01/09/2014
From:	Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
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ML14015A117	List: ML14015A096 ML14015A097 ML14015A098 ML14015A101 ML14015A103 ML14015A104 ML14015A105 ML14015A106 ML14015A107 ML14015A108 ML14015A111 ML14015A112 ML14015A129 ML14015A131 ML14015A132 ML14015A134 ML14015A135 ML14015A149 ML14015A150 ML14015A152 ML14015A245 ML14015A246 ML14015A250 ML14015A251 ML14015A252 ML14015A253 ML14015A254 ML14015A255 ML14015A283 ML14015A284 ML14015A285 ML14015A286 ML14015A290 ML14015A291 ML14015A292 RS-14-006, Clinton, Unit 1, Updated Licensing Basis Documents RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16 Chapter 3 - Design of Structures, Components, Equipment, and Systems, Figure 3.6-1. Sheet 95 of 111 Through Figure 3.6-1. Sheet 100 of 111 RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Appendix B - Construction Material Standards and Quality Control Procedures RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Appendix C - Computer Programs RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Appendix a - Glossary RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 1 - Introduction and General Description of Plant RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 10 - Steam and Power Conversion System RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 11 - Radioactive Waste Management RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 13 - Conduct of Operations RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 14 - Initial Test Program RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 15 - Accident Analysis RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 16 - Technical Specifications RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 17 - Quality Assurance RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 2 - Site Characteristics RS-14-006, Clinton, Unit 1, Updated Safety Analysis Report, Revision 16, Chapter 2 - Site Characteristics, Figure 2.1-1 Through Figure 2.4-18 ... further results
References
RS-14-006
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CPS/USAR CHAPTER 01 1-i REV. 13, JANUARY 2009 CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT TABLE OF CONTENTS PAGE 1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1-1

1.1 INTRODUCTION

1.1-1 1.1.1 Type of License Required 1.1-1 1.1.2 Identification of Applicant 1.1-1 1.1.3 Number of Plant Units 1.1-1 1.1.4 Description of Location 1.1-1 1.1.5 Type of Nuclear Steam Supply 1.1-1 1.1.6 Type of Containment 1.1-2 1.1.7 Core Thermal Power Levels 1.1-2 1.1.8 Scheduled Completion and Operation Dates 1.1-2 1.1.9 Organization of Contents 1.1-2 1.1.9.1 Subdivisions 1.1-2 1.1.9.2 Standard Format 1.1-2 1.1.9.3 References 1.1-2 1.1.9.4 Tables and Figures 1.1-3 1.1.9.5 Page Numbering 1.1-3 1.1.9.6 Revisions 1.1-3 1.1.9.7 Maintenance of the USAR 1.1-3 1.2 GENERAL PLANT DESCRIPTION 1.2-1 1.2.1 Principal Design Criteria 1.2-1 1.2.1.1 General Design Criteria 1.2-1 1.2.1.1.1 Power Generation Design Criteria 1.2-1 1.2.1.1.2 Safety Design Criteria 1.2-2 1.2.1.2 System Criteria 1.2-4 1.2.1.2.1 Nuclear System Criteria 1.2-4 1.2.1.2.2 Power Conversion Systems Criteria 1.2-5 1.2.1.2.3 Electrical Power Systems Criteria 1.2-5 1.2.1.2.4 Radwaste System Criteria 1.2-5 1.2.1.2.5 Auxiliary Systems Criteria 1.2-6 1.2.1.2.6 Shielding and Access Control Criterion 1.2-6 1.2.1.2.7 Nuclear Safety Systems and Engineered Safety Features Criteria 1.2-6 1.2.1.2.8 Process Control Systems Criteria 1.2-7 1.2.1.2.8.1 Nuclear System Process Control Criteria 1.2-7 1.2.1.2.8.2 Power Conversion Systems Process Control Criteria 1.2-7 1.2.1.2.8.3 Electrical Power System Process Control Criteria 1.2-7 1.2.2 Station Description 1.2-8 1.2.2.1 Site Characteristics 1.2-8 CPS/USAR TABLE OF CONTENTS (cont'd)

PAGE CHAPTER 01 1-ii REV. 13, JANUARY 2009 1.2.2.1.1 Location 1.2-8 1.2.2.1.2 Description of Station Environs 1.2-8 1.2.2.1.2.1 Meteorology 1.2-9 1.2.2.1.2.2 Hydrology 1.2-9 1.2.2.1.2.3 Geology 1.2-9 1.2.2.1.2.4 Seismology 1.2-10 1.2.2.1.3 Design Bases Dependent on Site Environs 1.2-11 1.2.2.2 General Arrangement of Structures and Equipment 1.2-11 1.2.2.3 Nuclear System 1.2-12 1.2.2.3.1 Reactor Core and Control Rods 1.2-12 1.2.2.3.2 Reactor Vessel and Internals 1.2-12 1.2.2.3.3 Reactor Recirculation System 1.2-13 1.2.2.3.4 Residual Heat Removal System 1.2-13 1.2.2.3.5 Reactor Water Cleanup System 1.2-13 1.2.2.3.6 Nuclear Leak Detection System 1.2-13 1.2.2.4 Nuclear Safety Systems and Engineered Safety Features 1.2-14 1.2.2.4.1 Reactor Protection System 1.2-14 1.2.2.4.2 Neutron Monitoring System 1.2-14 1.2.2.4.3 Control Rod Drive System 1.2-14 1.2.2.4.4 Control Rod Drive Housing Supports 1.2-14 1.2.2.4.5 Control Rod Velocity Limiter 1.2-15 1.2.2.4.6 Nuclear System Pressure Relief System 1.2-15 1.2.2.4.7 Reactor Core Isolation Cooling System 1.2-15 1.2.2.4.8 Emergency Core Cooling Systems 1.2-15 1.2.2.4.9 Containment Systems 1.2-16 1.2.2.4.9.1 Containment Functional Design 1.2-16 1.2.2.4.9.2 Residual Heat Removal System (Suppression Pool Cooling) 1.2-17 1.2.2.4.9.3 Residual Heat Removal System (Containment Spray) 1.2-17 1.2.2.4.9.4 Combustible Gas Control 1.2-18 1.2.2.4.10 Containment and Reactor Vessel Isolation Control System 1.2-18 1.2.2.4.10.1 Main Steamline Isolation Valves 1.2-18 1.2.2.4.10.2 Main Steamline Flow Restrictors 1.2-18 1.2.2.4.11 Process Radiation Monitoring System 1.2-19 1.2.2.4.11.1 Main Steamline Radiation Monitoring System 1.2-19 1.2.2.4.11.2 Containment Building Ventilation Radiation Monitoring System 1.2-19 1.2.2.4.11.3 Fuel Building Ventilation Radiation Monitoring System 1.2-19 1.2.2.4.12 Standby Gas Treatment System 1.2-19 1.2.2.4.13 Safety-Related Electrical Power Systems 1.2-19 1.2.2.4.14 Standby Liquid Control System 1.2-20 1.2.2.4.15 Safe Shutdown from Outside the Control Room 1.2-20 1.2.2.4.16 Main Steamline Isolation Valve Leakage Control System 1.2-20 1.2.2.4.17 Suppression Pool Makeup System 1.2-20 1.2.2.4.18 Control Room HVAC 1.2-20 1.2.2.4.19 Shutdown Service Water System 1.2-20 1.2.2.5 Power Conversion System 1.2-21 1.2.2.5.1 Turbine-Generator 1.2-21 1.2.2.5.2 Main Steam System 1.2-21 CPS/USAR TABLE OF CONTENTS (cont'd)

PAGE CHAPTER 01 1-iii REV. 13, JANUARY 2009 1.2.2.5.3 Main Condenser 1.2-21 1.2.2.5.4 Main Condenser Evacuation System 1.2-21 1.2.2.5.5 Turbine Gland Sealing System 1.2-21 1.2.2.5.6 Steam Bypass System and Pressure Control System 1.2-21 1.2.2.5.7 Circulating Water System 1.2-22 1.2.2.5.8 Condensate and Feedwater System 1.2-22 1.2.2.5.9 Condensate Cleanup System 1.2-22 1.2.2.6 Electrical Systems and Instrumentation and Control 1.2-22 1.2.2.6.1 Electrical Power Systems 1.2-22 1.2.2.6.1.1 138-kV Offsite Power System 1.2-23 1.2.2.6.1.2 345-kV Offsite Power System 1.2-23 1.2.2.6.1.3 Unit Auxiliary A-C Power System 1.2-23 1.2.2.6.1.4 Unit Class lE A-C Power System 1.2-23 1.2.2.6.1.5 Nuclear System Protection System Power System 1.2-23 1.2.2.6.1.6 Instrument Power System 1.2-23 1.2.2.6.1.7 Uninterruptible Power System 1.2-23 1.2.2.6.1.8 Unit Auxiliary D-C Power System 1.2-24 1.2.2.6.1.9 Unit Class lE D-C Power System 1.2-24 1.2.2.6.2 Nuclear System Process Control and Instrumentation 1.2-24 1.2.2.6.2.1 Rod Control and Information System 1.2-24 1.2.2.6.2.2 Recirculation Flow Control System 1.2-24 1.2.2.6.2.3 Neutron Monitoring System 1.2-24 1.2.2.6.2.4 Refueling Interlocks 1.2-25 1.2.2.6.2.5 Reactor Vessel Instrumentation 1.2-25 1.2.2.6.2.6 Process Computer System 1.2-25 1.2.2.6.3 Power Conversion Systems Process Control and Instrumentation 1.2-25 1.2.2.6.3.1 Pressure Regulator and Turbine-Generator Control 1.2-25 1.2.2.6.3.2 Feedwater Control System 1.2-25 1.2.2.7 Fuel Handling and Storage Systems 1.2-25 1.2.2.7.1 New and Spent Fuel Storage 1.2-25 1.2.2.7.2 Fuel Handling System 1.2-26 1.2.2.8 Cooling Water and Auxiliary Systems 1.2-26 1.2.2.8.1 Closed Cooling Water Systems 1.2-26 1.2.2.8.1.1 Component Cooling Water System 1.2-26 1.2.2.8.1.2 Turbine Building Closed Cooling Water System 1.2-27 1.2.2.8.2 Fuel Pool Cooling and Cleanup System 1.2-27 1.2.2.8.3 Plant Service Water System 1.2-27 1.2.2.8.4 Ultimate Heat Sink 1.2-27 1.2.2.8.5 Condensate Storage and Transfer System 1.2-27 1.2.2.8.6 Raw Water Treatment Plant and Makeup Water Treatment System 1.2-27 1.2.2.8.7 Potable and Sanitary Waste Water System 1.2-28 1.2.2.8.8 Plant Chilled Water Systems 1.2-28 1.2.2.8.9 Process Sampling System 1.2-28 1.2.2.8.10 Plant Equipment and Floor Drainage 1.2-28 1.2.2.8.11 Service and Instrument Air Systems 1.2-28 CPS/USAR CHAPTER 01 1-iv REV. 13, JANUARY 2009 1.2.2.8.12 Diesel Generator Fuel-Oil Storage and Transfer System 1.2-29 1.2.2.8.13 Auxiliary Steam Systems 1.2-29 1.2.2.8.14 Heating, Ventilation, and Air Conditioning/ Environmental Systems 1.2-29 1.2.2.8.15 Lighting Systems 1.2-30 1.2.2.8.16 Fire Protection System 1.2-30 1.2.2.8.17 Communication Systems 1.2-31 1.2.2.9 Radioactive Waste Systems 1.2-31 1.2.2.9.1 Gaseous Radwaste System 1.2-31 1.2.2.9.2 Liquid Radwaste System 1.2-31 1.2.2.9.3 Solid Radwaste System 1.2-32 1.2.2.10 Radiation Monitoring and Control 1.2-32 1.2.2.10.1 Process Radiation Monitoring 1.2-32 1.2.2.10.2 Area Radiation Monitors 1.2-32 1.2.2.10.3 Site Environs Radiation Monitors 1.2-32 1.2.2.11 Shielding 1.2-33 1.3 COMPARISON TABLES 1.3-1 1.3.1 Comparison with Similar Facility Designs 1.3-1 1.3.2 Comparison of Final and Preliminary Information 1.3-1 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 1.4.1 Applicant 1.4-1 1.4.2 Architect-Engineer - Sargent & Lundy 1.4-1 1.4.3 General Electric 1.4-1 1.4.3.1 Nuclear Steam Supply System 1.4-1 1.4.3.2 Turbine-Generator 1.4-1 1.4.4 Constructor 1.4-2 1.4.5 Technical Consultants 1.4-3 1.4.5.1 Dames & Moore 1.4-3 1.4.5.2 Hazleton Environmental Sciences Corporation 1.4-3 1.4.5.3 The Research Corporation of New England 1.4-4 1.4.5.4 Harza Engineering 1.4-4 1.4.5.5 Nuclear Services Corporation 1.4-4 1.4.5.6 Southwest Research Institute 1.4-4 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.5.1 Current Development Programs 1.5-1 1.5.1.1 Instrumentation for Vibration 1.5-1 1.5.1.2 Core Spray Distribution 1.5-1 1.5.1.3 Core Spray and Core Flooding Heat Transfer Effectiveness 1.5-1 1.5.1.4 Verification of Pressure Suppression Design 1.5-1 1.5.1.5 Boiling Transition Testing 1.5-3 CPS/USAR CHAPTER 01 1-v REV. 13, JANUARY 2009 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1.7-1 1.7.1 Electrical, Instrumentation, and Control Drawings 1.7-1 1.7.2 Piping and Instrumentation Diagrams, Mechanical, Physical, Electrical, and Other Drawings Used in the USAR 1.7-1 1.7.3 Other Detailed Information 1.7-1 1.7.4 Process Diagrams and Other Figures 1.7-1 1.8 CONFORMANCE TO NRC REGULATORY GUIDES 1.8-1 1.9 SYMBOLS USED IN ENGINEERING DRAWINGS 1.9-1 1.10 ACRONYMS 1.10-1 CPS/USAR CHAPTER 01 1-vi REV. 13, JANUARY 2009 CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANTS LIST OF TABLES NUMBER TITLE PAGE 1.3-1 Comparison of Nuclear System Design and Operating Characteristics 1.3-2 1.3-2 Comparison of Power Conversion System Design Characteristics 1.3-9 1.3-3 Comparison of Electrical Systems 1.3-11 1.3-4 Comparison of Containment Design Characteristics 1.3-12 1.3-5 Comparison of Structural Design Requirements 1.3-14 1.3-6 Comparison of Standby Gas Treatment Systems 1.3-15 1.3-7 Comparison of Radioactive Waste Management 1.3-16 1.3-8 Significant Design Changes From PSAR to FSAR 1.3-18 1.5-1 Summary of PSTF Tests 1.5-4 1.5-2 References 1.5-6 1.6-1 Referenced Reports 1.6-2 1.8-1 Conformance to IEEE Standards 1.8-2 CPS/USAR CHAPTER 01 1-vii REV. 13, JANUARY 2009 CHAPTER 1 - INTRODUCTION AND GENERAL DESCRIPTION OF PLANT LIST OF FIGURES NUMBER TITLE 1.1-1 Heat Balance at Rated Power 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 USAR. They are controlled by the Controlled Documents Program. 197R567 Piping and Instrument Symbols (General Electric) 209A4756 Logic Symbols (General Electric) 209A7367 Solid State Logic Symbols (General Electric) 921D280 Instrument Symbols (General Electric)

EC-110 Graphic Symbols for power and Control Circuits EC-111 Graphic Symbols for Key Diagrams of Auxiliaries M01-1101 Site Development M01-1102 Site Development M01-1103 Site Development M01-1105 General Arrangement - Basement Floor Plan M01-1106 General Arrangement - Grade Floor Plan El. 737'-0" M01-1107 General Arrangement - Mezzanine Floor Plan El. 762'-0" M01-1108 General Arrangement - Main Floor Plan M01-1109 General Arrangement - Miscellaneous Floor Plans M01-1110 General Arrangement - Sections "A-A" and "B-B" M01-1111 General Arrangement - Sections "C-C", "D-D" and "E-E" M01-1112 General Arrangement - Sections "F1-F1", "F2-F2" and "G-G" M01-1113 General Arrangement - Sections "H-H" and "J-J" M01-1114 General Arrangement - Section "K-K" M01-1115 General Arrangement - Roof Plan M01-1116 General Arrangement - Circulating Water Screen House M01-1119 Makeup Water Pump House and Storage Tank Grade Floor Plan M05-1000 P&ID Index M05-1001 P&ID Symbols - S&L

CPS/USAR CHAPTER 01 1.6-1 REV. 11, JANUARY 2005 1.6 MATERIAL INCORPORATED BY REFERENCE Table 1.6-1 is a list of GE topical reports and any other report or document which is incorporated in whole or in part by reference in this USAR and has been filed with the NRC. Additional documents which are referenced in this USAR are listed at the end of the sections in which they have been referenced.

CPS/USAR CHAPTER 01 1.6-2 REV. 11, JANUARY 2005 TABLE 1.6-1 REFERENCED REPORTS GENERAL ELECTRIC COMPANY REPORTS REPORT NUMBER TITLE REFERENCED IN USAR SECTION APED-4827 Maximum Two-Phase Blowdown from Pipes (April 1965) 6.2 APED-4986 Consequences of Operating Zircaloy-2 Clad Fuel Rods Above the Critical Heat Flux (October 1965 BWR 6 only) 4.2 APED-5286 Design Basis for Critical Heat Flux Condition in BWR's (September 1966) 1.5 APED-5458 Effectiveness of Core Standby Cooling Systems for General Electric Boiling Water Reactors (March 1968) 5.4 APED-5460 Design and Performance of General Electric BWR Jet Pumps (July 1968) 3.9 APED-5555 Impact Testing on Collet Assembly for Control Rod Drive Mechanism 7RDB144A (November

1967) 4.6 APED-5640 Xenon Considerations in Design of Large Boiling Water Reactors (June 1968) 4.1 4.3 APED-5652 Stability and Dynamic Performance of the General Electric Boiling Water Reactor (April

1969) 4.1 APED-5706 In-Core Neutron Monitoring System for General Electric Boiling Water Reactors (November 1968, Revised April 1969) 7.6 7.7 APED-5736 Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards (April 1969)

Appendix 15A APED-5750 Design and Performance of General Electric Boiling Water Reactor Main Steamline Isolation Valves (March 1969) 5.4 APED-5756 Analytical Methods for Evaluating the Radiological Aspects of the General Electric

Boiling Reactor (March 1969) 15.4, 15.7 CPS/USAR TABLE 1.6-1 (Cont'd) CHAPTER 01 1.6-3 REV. 11, JANUARY 2005 REPORT NUMBER TITLE REFERENCED IN USAR SECTION GEAP-4616 Two-Phase Pressure Drop in Straight Pipes and Channels; Water-Steam Mixtures at 600 to 1400 psia (May 1964) 4.4 GEAP-10546 Theory Report for Creep-Plast Computer Program (January 1972) 4.1 GEAP-13112 Thermal Response and Cladding Performance of an Internally Pressurized Zircaloy-Clad, Simulated BWR Bundle Cooled by Spray Under Loss-of-Coolant Conditions (April 1971) 4.2 KAPL-2170 Hydrodynamic Stability of a Boiling Channel (October 1961) 4.4 KAPL-2208 Hydrodynamic Stability of a Boiling Channel, Part 2 (April 1962) 4.4 KAPL-2290 Hydrodynamic Stability of a Boiling Channel, Part 3 (June 1963) 4.4 KAPL-3070 Hydrodynamic Stability of a Boiling Channel, Part 4 (August 1964) 4.4 KAPL-3072 Reactivity Stability of a Boiling Reactor, Part 1 (September 1964) 4.4 KAPL-3093 Reactivity Stability of a Boiling Reactor, Part 2 (March 1965) 4.4 NEDC-20944 Peachbottom Atomic Power Station Units 2 and 3, Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core

Vibration (December 1975) 4.4 NEDE-10313 PDA-Pipe Dynamic Analysis Program for Pipe Rupture Movement (Proprietary (Filing) 3.6, 3.9 NEDE-11146 Design Basis for New Gas System (July 1971)(Company Proprietary) 11.3 NEDE-20386 Fuel Channel Deflections 4.2 NEDE-20943 Urania-Gadolinia Nuclear Fuel Physical and Irradiation Characteristics and Material

Properties (January 1977) 4.2 CPS/USAR TABLE 1.6-1 (Cont'd) CHAPTER 01 1.6-4 REV. 11, JANUARY 2005 REPORT NUMBER TITLE REFERENCED IN USAR SECTION NEDE-20944-P BWR/4 and BWR/5 Fuel Design (October 1976) (only BWR/4&5) 4.2, 4.3, 4.4 and 4.6 NEDE-20944-1P BWR/4 and BWR/5 Fuel Design (Amendment

1) (January 1977) (only BWR/4&5) 4.2, 4.3, 4.4 and 4.6 NEDE-21156 Supplemental Information for Plant Modification to Eliminate Significant In-Core

Vibration (January 1976) 4.4 NEDE-21175-P BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA)

Loadings (November 1976) 3.9 NEDE-21354-P BWR Fuel Channel Mechanical Design and Deflection (September 1976) 3.9 NEDE-23014 Hex 01 User's Manual (July 1976) 15.2 NEDE-23542-P and NEDO-23542 Fuel Assembly Evaluation of Shipping and Handling Loadings (March 1977) (Proprietary and Non-proprietary Versions) 4.2 NEDE-24154-P and NEDO-24154 Safety Evaluation for General Electric Topical Report: Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors 15.0 15.1 Appendix 15B Appendix 15D NEDM-10735 Densification Considerations in BWR Fuel Design and Performance (December 1972) 4.2 NEDO-10173 Current State of Knowledge, High Performance BWR Zircaloy-Clad UO2 Fuel (May 1973) 4.2 11.1 NEDO-10174 Consequences of a Postulated Fuel Blockage Incident in a Boiling Water Reactor (May 1970) 4.2 NEDO-10722A Core Flow Distribution in a Modern Boiling Water Reactor as Measured in Monticello (August 1976) 4.4 CPS/USAR TABLE 1.6-1 (Cont'd) CHAPTER 01 1.6-5 REV. 11, JANUARY 2005 REPORT NUMBER TITLE REFERENCED IN USAR SECTION NEDO-10320 The General Electric Pressure Suppression Containment Analytical Model (April 1971)

Supplement 1 (May 1971) 6.2 NEDO-10329 Loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors (April 1971)

Supplement 1 (April 1971) Addenda (May 1971) 4.3 NEDO-10349 Analysis of Anticipated Transients Without Scram (March 1971) 15.8 NEDO-10466A Power Generation Control Complex Design Criteria and Safety Evaluation (February 1979) 6.4, 9.5 NEDO-10505 Experience with BWR Fuel Through September 1971 (May 1972) 4.2, 11.1 NEDO-10527 Rod Drop Accident Analysis for Large Boiling Water Reactors (March 1972) Supplement 1 (July 1972) Supplement 2 (January 1973) 4.3, 15.4 NEDO-10585 Behavior of Iodine in Reactor Water During Plant Shutdown and Startup (August 1972) 15.6 NEDO-10602 Testing of Improved Jet Pumps for the BWR/6 Nuclear System (June 1972) 3.9 NEDO-10734 A General Justification for Classification of Effluent Treatment System Equipment as Group D (February 1973) 11.3 NEDO-10739 Methods for Calculating Safe Test Intervals And Allowable Repair Times for Engineered

Safeguard Systems (January 1973) 6.3, 15A NEDO-10751 Experimental and Operational Confirmation of Off-Gas System Design Parameters (January

1973) (Company Proprietary) 11.3 NEDO-10801 Modeling the BWR/6 Loss-of-Coolant Accident: Core Spray and Bottom Flooding

Heat Transfer Effectiveness (March 1973) 1.5 CPS/USAR TABLE 1.6-1 (Cont'd) CHAPTER 01 1.6-6 REV. 11, JANUARY 2005 REPORT NUMBER TITLE REFERENCED IN USAR SECTION NEDO-10802 Analytical Methods of Plant Transient Evaluations for General Electric Boiling Water

Reactor (February 1973) 4.4, 5.2, 15.1 NEDO-10846 BWR Core Spray Distribution (April 1973) 1.5 NEDO-10899 Chloride Control in BWR Coolants (June 1973) 5.2 NEDO-10905 High Pressure Core Spray System Power Supply Unit (May 1973) 8.1 NEDO-10958 General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and

Design Application (November 1973) 4.3, 4.2, 15.0 NEDO-10958-A General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application (January 1977) 1.5, 4.4, 15.4, NEDO-20231 Emergency Core Cooling Tests of an Internally Pressurized, Zircaloy-Clad, 8 x 8

Simulated BWR Fuel Bundle (December 1973) 1.5 NEDO-20340 Process Computer Performance Evaluation Accuracy (June 1974) 4.3 NEDO-20360 General Electric Boiling Water Reactor Generic Reload Application for 8 x 8 Fuel (May 1975) 4.2, 15.4 NEDO-20360-IP General Electric Boiling Water Reactor Generic Reload Application for 8 x 8 Fuel (March 1976) 4.2 NEDO-20377 8 x 8 Fuel Bundle Development Support (February 1975) 4.2 NEDO-20533 The General Electric Mark III Pressure Suppression Containment System Analytical

Model (June 1974) 1.5 NEDO-20566 and NEDE-20566-P General Electric Company Model for Loss-of-Coolant Accident Analysis in Accordance with 10 CFR 50, Appendix K (January 1976) 3.9 CPS/USAR TABLE 1.6-1 (Cont'd) CHAPTER 01 1.6-7 REV. 11, JANUARY 2005 REPORT NUMBER TITLE REFERENCED IN USAR SECTION NEDO-20605 and NEDE-20606 Creep Collapse Analysis of BWR Fuel Using Safe Collapse Model (August 1974);

(Nonproprietary and Proprietary Versions) 4.2 NEDO-20626 Studies of BWR Designs for Mitigation of Anticipated Transients Without Scrams (October 1974) 15.8 NEDO-20626-1 Studies of BWR Designs for Mitigation of Anticipated Transients Without Scrams (June 1975) 15.8 NEDO-20626-2 Studies of BWR Designs for Mitigation of Anticipated Transients Without Scrams (July

1975) 15.8 NEDO-20631 Mechanical Property Surveillance of Reactor Pressure Vessels for General Electric BWR/6

Plants (March 1975) 5.3 NEDO-20913 Lattice Physics Methods (June 1975) 4.3 NEDO-20922 Experience With BWR Fuel Through September 1974 (June 1975) 4.2, 11.1 NEDO-20939 Lattice Physics Methods Verification (August 1975) 4.3 NEDO-20943 Urania-Gadolinia Nuclear Fuel Physical and Material Properties (January 1977) 4.2 NEDO-20944 BWR/4 and BWR/5 Fuel Design (October 1976) (Nonproprietary Versions) 4.1, 4.3 NEDE-20944 BWR/4 and BWR/5 Fuel Design (October 1976) (Proprietary Version) 4.1, 1.3 NEDO-20946 BWR Simulator Methods Verification (May 1976) 4.3 NEDO-20948-P BWR/6 Fuel Design (June 1976) 4.2 NEDO-20953 Three-Dimensional Boiling Water Reactor Core Simulator (May 1976) 15.4 CPS/USAR TABLE 1.6-1 (CONT'D) CHAPTER 01 1.6-8 REV. 12, JANUARY 2007 REPORT NUMBER TITLE REFERENCED IN USAR SECTION NEDO-20964 Generation of Void and Doppler Reactivity Feedback for Application to BWR Plant Transient Analysis (August 1975) 4.3 NEDO-21142 Realistic Accident Analysis for General Electric Boiling Water Reactor - The RELAC Code and User's Guide, (December 1977) 15.6, 15.7 NEDO-21159 Airborne Release from BWR's for Environment Impact Evaluations (March

1976) 11.1 NEDO-21231 Banked Position Withdrawal Sequence (September 1976) 4.3 NEDO-21291 Group Notch Mode of the RSCS for Cooper (June 1976) 15.4 NEDO-21506 Stability and Dynamic Performance of the General Electric Boiling Water Reactor (January 1977) 4.4 NEDO-26453 3D BWR Core Simulator (May 1976) Oyster Creek Station, FSAR Amendment 10 4.3, 1.5 NEDO-21660 Experience with BWR Fuel Through December 1976 (July 1977) 4.2 NEDE-24011-P-A General Electric Standard Application for Reactor Fuel (latest approved revision) 4.2, 4.3, 4.4, Appendix 15D NEDE-30130-P Steady State Nuclear Methods (April 1985) Appendix 15D OTHER REFERENCED REPORTS AE-RTL-788 Void Measurements in the Region of Sub- cooled and Low Quality Boiling (April 1966) 4.4 ANL-5621 Boiling Density in Vertical Rectangular Multichannel Sections with Natural Circulation (November 1956) 4.4 CPS/USAR TABLE 1.6-1 (CONT'D) CHAPTER 01 1.6-9 REV. 11, JANUARY 2005 REPORT NUMBER TITLE REFERENCED IN USAR SECTION ANL-5552 The Effect of Pressure on Boiling Density in Multiple Rectangular Channel (February

1956) 4.4 ANL-6385 Power-to-Void Transfer Functions (July 1961) 4.4 BHR/DER 70-1 Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor (March 1970) 11.1 BMI-1163 Vapor Formation and Behavior in Boiling Heat Transfer (February 1957) 4.4 CF 59-6-47 (ORNL) Removal of Fission Product Gases from Reactor Off-Gas Streams by Adsorption 11.3 IDO-ITR-105 The Response of Waterlogged UO2 Fuel Rods to Power Bursts (April 1969) 4.2 IN-ITR-111 The Effects of Cladding Material and Heat Treatment on the Response of Water-logged UO2 Fuel Rods to Power Bursts (January

1970) 4.2 RE-S-76-170 Light Water Reactor Fuel Behavior Program Description; RIA Fuel Behavior Experiment Requirements (September 1976) 4.2 STL-372-38 Kinetic Studies of Heterogeneous Water Reactors (April 1966) 4.4 TID-4500 Relap 3 - A Computer Program for Reactor Blowdown Analysis IN-1321 (June 1970) 3.6 TID-7672 ANS Topical Meeting, Nuclear Performance of Power Reactors (September 1976) 4.3 UCRL-50451 Improving Availability and Readiness of Field Equipment Through Periodic Inspection, P. 10 (July 16, 1968) 16.3 WACP-6065 Melting Point of Irradiated Uranium Dioxide (February 1965) 4.2 WAPD-BT-19 A Method of Predicting Steady-State Boiling Vapor Fractions in Reactor Coolant Channels (June 1960) 4.4 CPS/USAR TABLE 1.6-1 (Cont'd) CHAPTER 01 1.6-10 REV. 11, JANUARY 2005 REPORT NUMBER TITLE REFERENCED IN USAR SECTION WAPD-TM-283 Effects of High Burnup on Zircaloy-Clad Bulk UO2 Plate Fuel Element Samples (September

1962) 4.2 WAPD-TM-416 WIGLE - A Program for the Solution of the Two-Group Space-Time Diffusion in Slab Geometry (1964) 4.3 WAPD-TM-629 Irradiation Behavior of Zircaloy-Clad Fuel Rods Containing Dished End UO2 Pellets (July 1967) 4.2 CPS/USAR CHAPTER 01 1.7-1 REV. 11, JANUARY 2005 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1.7.1 Electrical, Instrumentation, and Control Drawings Electrical, instrumentation, and control drawings were provided to the NRC during initial licensing activities. These drawings were considered necessary to evaluate the safety-related features in Chapters 7 and 8. These drawings are not required to be included in the USAR per Generic Letter 81-06, Question/Response C.1. 1.7.2 Piping and Instrumentation Diagrams (P&IDs), Mechanical, Physical, Electrical, and Other Drawings Used in the USAR The plant drawings cited in each Chapter of Appendix of the USAR are listed in the Table of Contents for that Chapter or Appendix. 1.7.3 Other Detailed Information No specific request for data has been received from the NRC; therefore, no information is supplied for this subsection. 1.7.4 Process Diagrams And Other Figures Current design basis documents and drawings should be referred to when reviewing and evaluating the design..

CPS/USAR CHAPTER 01 1.8-1 REV. 11, JANUARY 2005 1.8 CONFORMANCE TO NRC REGULATORY GUIDES The purpose of this section is to indicate the conformance of the Clinton Power Station design with Regulatory Guides issued by the NRC. On the following pages each Regulatory Guide is identified by number, revision and title. The project position toward the Regulatory Guide is also indicated, accompanied by the appropriate USAR reference section to which the Regulatory Guide applies. If the project position indicates compliance, no further discussion is provided. Exceptions taken to the Regulatory Guide are identified and discussed in this section. Clarifications to compliance positions are also provided in this section when necessary. Conformance to IEEE Standards - Also included in this section as Table 1.8-1 is a list of IEEE Standards which were considered for the design, construction and operation of the Clinton Power Station. These specific revisions of the Standards apply to the CPS-USAR except as otherwise noted in the USAR text.

CPS/USAR CHAPTER 01 1.8-2 REV. 11, JANUARY 2005 Table 1.8-1 CONFORMANCE TO IEEE STANDARDS IEEE Standard 85 (1973) Test Procedure for Airborne Sound Measurement on Rotating Electrical Machinery IEEE Standard 112A (1964) Test Procedure for Polyphase Induction Motors and Generators IEEE Standard 275(1981) Recommended Practice for Thermal Evaluation of Insulation Systems for AC Electric Machinery Employing Form-Wound Pre-Insulated Stator Coils Machines Rated 6900 V and BelowIEEE Standard 279 (1971) Criteria for Protection Systems for Nuclear Power Generating Stations IEEE Standard 308(1974) Criteria for Class 1E Power Systems for Nuclear Power Generating Stations IEEE Standard 317(1976) Electrical Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations IEEE Standard 323(1974) Qualifying Class 1E Equipment for Nuclear Power Generating Stations IEEE Standard 334(1974) Standard for Type Test of Continuous Duty Class 1E Motors for Nuclear Power Generating Stations IEEE Standard 336(1971) Installation, Inspection, and Testing Requirements for Instrumentation and Electrical Equipment During the Construction of Nuclear Power Generating Stations IEEE Standard 338(1977) Standard Criteria for the Periodic Testing of Nuclear Power Generating Station Class 1E Power and Protection Systems IEEE Standard 344(1975) Recommended Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations IEEE Standard 379(1972) Trial-Use Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection Systems IEEE Standard 381(1977) Standard Criteria for Type-Test of Class 1E Modules Used in Nuclear Power Generating Stations IEEE Standard 382(1972) Trial-Use Guide for the Type-Test of Class 1E Electrical Valve Operators for Nuclear Power Generating Stations CPS/USAR CHAPTER 01 1.8-3 REV. 14, JANUARY 2011 Table 1.8-1 (Cont'd) Conformance to IEEE Standards IEEE Standard 383(1974) Standard for Type Test of Class 1E Electrical Cables, Field Splices, and Connections for Nuclear Power Generating Stations IEEE Standard 384(1974) Trial-Use Standard Criteria for Separation of Class 1E Equipment and Circuits IEEE Standard 387(1977) Trial-Use Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations IEEE Standard 415(1976) Planning of Pre-Operational Testing Programs for Class 1E Power Systems for Nuclear Power Generating Stations IEEE Standard 450(1995) IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Station Applications IEEE Standard 484(1975) Recommended Practice for Installation Design and Installation of Large Lead Storage Batteries for Generating Stations and Substations IEEE Standard 622(1979) IEEE Recommended Practice for the Design and Installation of Electric Pipe Heating Systems for Nuclear Power

Generating Stations IEEE Standard 627(1980) IEEE Standard for Design Qualification of Safety Systems Equipment Used in Trial-Use Criteria for Nuclear Power

Generating Stations IEEE Standard 634(1978) IEEE Standard Cable Penetration Fire Stop Qualification Test

- (see also NEDO-10905)

CPS/USAR CHAPTER 01 1.8-4 REV. 12, JANUARY 2007 Regulatory Guide 1.1, Rev. 0 (December 1970)

Net Positive Suction Head For Emergency Core Cooling and Containment Heat Removal System Pumps Project Position - Comply USAR Subsection - 6.3.2.2 Regulatory Guide 1.2, Rev. 0 (December 1970)

Thermal Shock to Reactor Pressure Vessels Project Position - Comply. USAR Subsection - 5.3.3 Regulatory Guide 1.3, Rev. 2 (June 1974)

Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors Project Position - Partially Comply; for Alternative Source Terms, Regulatory Guide 1.183 is also utilized. USAR Subsections - 6.4.2, 7.6.2.12.5, 9.3.7, 15.6, E3.8.1.2 Regulatory Guide 1.4, Rev. 2 (June 1974)

Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors Project Position - Not applicable to BWRs.

CPS/USAR CHAPTER 01 1.8-5 REV. 12, JANUARY 2007 Regulatory Guide 1.5, Rev. 0 (March 1971)

Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors Project Position - Partially Comply; for Alternative Source Terms, Regulatory Guide 1.183 is also utilized. USAR Subsection - 15.6.4 Regulatory Guide 1.6, Rev. 0 (March 1971)

Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems Project Position - Comply USAR Subsections - 7.3.2.1.2.1.1, 7.3.2.3.2.1.1, 7.3.2.20.2.1.1, 8.3.1.2.2, 8.3.2.2.2 Regulatory Guide 1.7, Rev. 2 (November 1978)

Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident Project Position - The project complies with the requirements of Regulatory Guide 1.7 with the following clarification:

The discussion contained in Regulatory Guide 1.7 states that the NRC has concluded that a lower flammability limit of 4 volume percent hydrogen in air or steam-air mixtures is well established and is adequately conservative. Regulatory Position C.6 states that materials which would yield hydrogen gas due to corrosion from ECCS or containment spray should be identified and their use should be limited as much as practical. In USAR Section 6.2.5 regarding the Combustible Gas Control System, hydrogen production due to corrosion following a LOCA has been evaluated based upon the analyses performed considering the amount of hydrogen producing materials allowed in containment by design. The specific quantities of hydrogen producing materials used in evaluating CPS conformance to Regulatory Guide 1.7 are listed in Table 6.2-49, and the resulting hydrogen concentrations as a function of time are exhibited in Figures 6.2-130a and 6.2-130b. CPS procedures and design baseline documentation ensure the limitation of postulated post-LOCA hydrogen concentrations inside containment and drywell in accordance with Regulatory Guide 1.7. USAR Subsections - 6.2.5, 9.4.7.2 CPS/USAR CHAPTER 01 1.8-6 REV. 11, JANUARY 2005 Regulatory Guide 1.8, Proposed Rev. 2 (February 1979)

Personnel Selection and Training Project Position - The project complies with the requirements of Regulatory Guide 1.8 with the following clarifications and exceptions: 1.

Reference:

Paragraph C.3.a, C.3.b, and C.3.c - Exception is taken to the requirement of minimum qualifications for individuals that direct or supervise the conduct of individual preoperational tests and startup tests, and individuals who review and approve preoperational and startup test procedures or results.

Instead, the following program will be used to qualify individuals: In order to ensure that the various Startup Group activities are performed by qualified personnel, three levels of qualifications are established: a. Level I

b. Level II

c. Level III The qualification levels required to perform specific startup organization activities is in accordance with the following table:

NOTE 1. The checkout and initial operation (C&IO) phase is a period during which checkout and testing is completed which is prerequisite to subsequent preoperational or acceptance tests. 2. The preoperational phase is a period during which preoperational test procedures (PTP's) and acceptance test procedures (ATP's) are performed. These tests, in general, are conducted on an integrated system or subsystem basis to verify that systems are capable of operating in a safe and efficient manner compatible with system design bases. By definition, "preoperational tests" are performed on nuclear safety-related systems. 3. The startup phase is that period beginning with preparations for fuel loading and extending through warranty tests.Startup test procedures (STP's) are conducted to verify the performance of equipment under actual operating conditions.

CPS/USAR CHAPTER 01 1.8-7 REV. 11, JANUARY 2005 Regulatory Guide 1.8, Proposed Rev. 2 (February 1979) (Cont'd) Minimum Qualification Level I II III Approve, checkout and initial operation test procedures (C&IO) X Approve acceptance test procedures (ATP's) X Approve preoperational test procedures (PTP's) X Approve startup test procedures (STP's) X

• Direct or supervise conduct of C&IO tests X

• Direct or supervise conduct of ATP's X *Direct or supervise conduct of PTP's X *Direct or supervise conduct of STP's X Evaluate test results of C&IO tests prerequisite to ATP X Evaluate test results of ATP's X Evaluate test results of PTP's X Evaluate test results of STP's X Supervision of test program X Certification of personnel X Evaluate test results of C&IO test prerequisite to PTP's X The process by which individuals are evaluated and certified to their appropriate qualification level, is as follows: a. The person being certified completes a reading list consisting of quality and work related documents as specified in the CPS Startup Manual. b. A resume or equivalent background information concerning education and experience is gathered. c. Prior to certification, the candidate performs work under the direction of certified personnel. The candidate's work qualities are documented and evaluated by the

Level III. d. A Level III will interview the candidate, discussing selected items from the reading list, past work experience applicabibilities to the present job, and considers the candidate's performance to item c above. The specified Level III will consider the individual for certification to one of three qualification levels.

• When an inspection or test requires implementation by a team or group, personnel not meeting the minimum qualification may be used for data taking assignments or equipment operation provided they are supervised by a qualified individual.

CPS/USAR CHAPTER 01 1.8-8 REV. 11, JANUARY 2005 Regulatory Guide 1.8, Proposed Rev. 2 (February 1979) (Cont'd) Delineated below are the requirements used when assigning a level of capability to an individual. The specified requirements are not to be treated as absolute. Other factors such as past performance or proficiency testing may be used to provide assurance that a person can competently perform an assigned task. When these requirements are waived, the basis for the waiver will be documented.

Level I High school graduate, plus one year of commensurate experience in construction, preoperational, startup or operational testing activities.

Level II (a) Graduate of four year accredited engineering or science college or university, plus two years of commensurate experience in construction, preoperational, startup, and/or operational testing activities. At least one year should be associated with nuclear facilities or if not, the individual should have training sufficient to acquaint him thoroughly with the safety aspects of a nuclear power plant. (b) High school graduate, plus four years of commensurate experience in construction, preoperational, startup, and/or operational testing activities in fossil or nuclear power plants, heavy industrial, or other similar equipment or facilities.

At least one year should be associated with nuclear facilities or if not, the individual should have training sufficient to acquaint him thoroughly with the safety aspects of a nuclear power plant.

Level III (a) Graduate of a four year accredited engineering or science college or university plus five years of experience in construction, startup, and/or operational testing activities. At least two years of this experience should be associated with preoperational and/or startup testing in nuclear facilities; or if not, the individual should have training sufficient to acquaint him thoroughly with the safety aspects of a nuclear power plant. (b) High school graduate, plus ten years of experience in testing, maintenance, or operational activities in nuclear or fossil power plants, heavy industrial, or other similar equipment or facilities. Five years of this experience should be associated with construction, preoperational, startup or operational testing and at least two of the five should be in nuclear facilities; or if not, the individual should have training sufficient to acquaint him thoroughly with the safety aspects of a nuclear power plant.

NOTE The word commensurate as used in the description of Level I and II is defined as: The knowledge and skills acquired through past experience corresponds in the same relative proportions to the knowledge and skill requirements of the tasks to CPS/USAR CHAPTER 01 1.8-9 REV. 11, JANUARY 2005 Regulatory Guide 1.8, Proposed Rev. 2 (February 1979) (Cont'd) be assigned to that test engineer. In other words, a person who had two years experience doing only electrical testing activities would not be assigned tasks of a mechanical or fluid nature. However, an engineer with two years experience would have gathered sufficient knowledge and skills to be assigned to an HVAC system startup. HVAC balancing would require additional experience. 2.

Reference:

Paragraph C.3.c - Clarification is provided concerning the qualifications of reviewers of preoperational and startup test procedures. The qualifications of reviewers and approvers is differentiated. The A/E or a consultant may be utilized to review portions of procedures. Furthermore, CPS staff personnel, who do not possess overall qualifications, may be utilized to review specific procedures if they are qualified in the particular area covered by the procedure. 3.

Reference:

Paragraph C.8 - Exception is taken to the requirement that non-licensed operators must have one year of power plant experience, 6 months of which must be at the assigned facility. This requirement would require the facility to carry at least one additional employee in an "in training" status since the auxiliary operator position is intended to be an entry level job. It is also believed that flexibility is lost when job vacancies are created. It is believed that there are numerous power plant duties having no safety significance that can be performed by an operator in an entry level position. This individual can receive on-the-job and formal training while contributing to plant productivity. 4.

Reference:

ANSI/ANS-3.1-1978, Paragraph 4.6.1 - The engineer in charge of technical support may not possess all specified qualifications. However, adequately qualified personnel will be available in the company engineering staff or through consultants to assist the engineer in charge. 5.

Reference:

ANSI/ANS-3.1-1978, Paragraph 4.7 - Independent review is provided by the Nuclear Safety Review Board (NSRB). The NSRB is a standing committee. Since the NSRB is a standing committee, paragraph 4.7.1 is not applicable to its members. 6.

Reference:

ANSI/ANS-3.1-1978, Paragraph 4.7 Exception is taken to the qualification requirements for the staff specialists for the review committee. The review committee is required to contain personnel with expertise in all of the appropriate areas. When such expertise is not available on the review committee unit, the expertise will be available in the company or from outside consultants. 7.

Reference:

ANSI/ANS-3.1-1978, Section 5.2 and Section 5.5 - Exception is taken to Section 5.2 "Training of Personnel to be Licensed by the NRC" and Section 5.5 "Operator Retraining and Replacement Training." The Clinton Power Station Licensed Operator Initial and Continuing Training Programs are accredited by the Institute of Nuclear Power Operations (INPO) National Academy for Nuclear Training. These programs are based on a systems approach to training (SAT) and are accredited by the National Nuclear CPS/USAR CHAPTER 01 1.8-10 REV. 11, JANUARY 2005 Regulatory Guide 1.8, Proposed Rev. 2 (February 1979) (Cont'd) Accrediting Board. These programs shall meet the requirements of 10CFR55, "Operators' Licenses." 8.

Reference:

ANSI/ANS-3.1-1978, Section 4, "Qualifications," Paragraph 4.2.2, "Operations Manager' - Exception is taken to the requirement "the Operations Manager shall hold a Senior Reactor Operator's License." The CPS Technical Specifications require "The Operations Manager or at least one Operations Middle Manager shall hold an SRO license for Clinton Power Station." This is consistent with ANSI/ANS-3.1-1978. USAR Subsection: 12.5.1, 13.1.3.1, 13.2.1, 13.2.1.1, 14.2.1.5, 14.2.2.5, TS 5.3.1, and ORM 6.4.1.

CPS/USAR CHAPTER 01 1.8-11 REV. 11, JANUARY 2005 Regulatory Guide 1.9, Rev. 2 (December 1979), and Rev. 3 Selection, Design and Qualification of Diesel-Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants Project Position - The project complies with NRC Regulatory Guide 1.9, Rev. 2 (December 1979), and with selected portions of Rev. 3, as described below: (1)

Reference:

Paragraphs C.1 and C.2 - Based upon updated diesel generator loading calculations, the loading on the diesel generators could be higher than the continuous rating of the diesel generator. Paragraph C.1 states that at the construction permit stage of design, the sum of the estimated loads needed at any one time is to be less than the continuous rating of the diesel generator. Paragraph C.2 also specifies that during the operating license stage of review, when a more accurate estimate of the safety loads is possible, a somewhat less conservative approach is permitted. Based on this, CPS uses the 2000-hour rating for illustrating the ability of the diesel to handle the maximum coincident loading expected following an accident. (2)

Reference:

Paragraph C.7 of Rev. 2 (December 1979)- Isolation valves are not included on instrument sensing lines, so that calibration of instrument sensors cannot be accomplished without disconnecting the sensor from the

sensed variable. (3)

Reference:

Section C, Paragraphs 2.2.2, 2.2.8, 2.2.9, and 2.2.10 of Reg. Guide 1.9, Rev. 3, establish the diesel-generator acceptance criteria and loading requirements for the load-run test, the full-load rejection test, the endurance and margin test, and the hot restart test. (The above referenced paragraphs of Reg. Guide 1.9, Rev. 3, were approved for use at CPS via Operating License Amendment 118.) (4)

Reference:

The General Electric HPCS system power supply unit Licensing Topical Report, NEDO 10905, gives the starting and accelerating charactersitics of the diesel-generator set with the various loads in the proper sequence. Although the voltage and frequency characteristics do not meet NRC Regulatory Guide 1.9, justification for this is given because of the unique requirements of the system. The HPCS diesel generator is unique in that its load is composed predominantly of one large motor whose horse power is approximately the same as the diesel-engine. USAR Subsection - 8.3.1.1.2, 8.3.1.1.2.1, 8.3.1.2.2 CPS/USAR CHAPTER 01 1.8-12 REV. 11, JANUARY 2005 Regulatory Guide 1.10, Rev. 1 (January 1973)

Mechanical (Cadweld) Splices in Reinforcing Bars of Category I Concrete Structures Project Position - The Project complies with NRC Regulatory Guide 1.10 with the following exceptions: (1)

Reference:

Paragraph C.1 - Each operator prepared two qualification splices for each of the positions (e.g., vertical, horizontal, diagonal) using the largest bar size for that position. (2)

Reference:

Paragraph C.1 - Operator requalification was necessary if (1) the specified splice position has not been used for a period of three months or more, (2) completed splices consistently failed to pass visual inspection or tensile test requirements, or (3) reason existed to question operator's ability. (3)

Reference:

Paragraph C.3.a - Testing complied with ASTM A370-75.

(4)

Reference:

Paragraph C.3 - Rebar detail drawings are prepared showing the location of all reinforcing bar lapped splices or cadwelds. If during construction the location of the splices or cadwelds differs from the location shown on the detail drawing by more than the specification tolerances, a Field Change Request (FCR) or Non-Conformance Report (NCR) is issued by the Constructor. When approval of the FCR or NCR is obtained, the number of the FCR or NCR is

posted against the drawing and becomes part of the permanent plant records. (5)

Reference:

Paragraph C.5.a - Procedure for Substandard Tensile Test Results were as follows: a. If any production or sister splice used for testing fails to meet the strength requirements (125% of minimum yield strength specified in ASTM A615-75) and failure occurs in the bar, the failure shall be reported to the Consulting Engineers. b. If any production splice used for testing fails to meet the strength requirements (125% of minimum yield strength specified in ASTM A615-

75) and failure did not occur in the bar, the adjacent production splices on each side of the failed splice shall be tested. If any sister splice used for testing fails to meet the strength requirements and failure did not occur in the bar, two additional sister splices shall be tested. If either of these retests fail to meet the strength requirements, splicing shall be halted.

Splicing shall not be resumed until the cause of failures has been corrected and resolved by the Contractor to the satisfaction of the Consulting Engineer. USAR Reference - Appendix B CPS/USAR CHAPTER 01 1.8-13 REV. 11, JANUARY 2005 Regulatory Guide 1.11, Rev. 0 (March, 1971)

Instrument Lines Penetrating Primary Reactor Containment Project Position - Comply with the following exception:

Reference:

Paragraph C.1.c(2) - Self-actuating Excess Flow Check Valves (EFCV's) that are installed in low pressure instrument sensing lines (i.e., lines that sense, drywell pressure, containment pressure, suppression pool level, and ventilation system pressure) are not designed to close if the instrument line integrity outside containment is lost during normal reactor operation. During normal reactor operation, there exists a small difference in atmospheric pressure between the drywell or containment buildings and the secondary containment, the building where all EFCV's are located. If an instrument line outside containment ruptures during normal reactor operation, there may be insufficient differential pressure to actuate the EFCV. However, since there is negligible radiological source term available for release from inside the drywell or containment during normal reactor operation, the safety consequence of a low pressure instrument sensing line failure is considered to be insignificant. The offsite radiological exposure from a single failure of an EFCV will remain substantially below the guidelines of 10CFR100 and the integrity and functional performance of the secondary containment and its associated standby gas

treatment system (SGTS) will be maintained. USAR Subsections - 6.2.4, 6.2.6.3, Table 6.2-47, 7.1.2.6.3, 7.3.2.2.2.1.1, 7.4.2.1.2.1.2 Regulatory Guide 1.12, Rev. 1 (April 1974)

Instrumentation for Earthquakes Project Position - Comply with the following exception:

Reference:

Paragraph C.4.b - Frequency range required for mechanical response spectrum recorder is minimum of 1 to 30 Hz. One of the recorders (passive) utilized at CPS is a 2-25.4 Hz recorder. USAR Subsection - 3.7.4 Regulatory Guide 1.13, Rev. 1 (December 1975)

Spent Fuel Storage Facility Design Basis Project Position - Comply USAR Section/Subsections - 3.1.2, 9.1.2, 9.1.3 CPS/USAR CHAPTER 01 1.8-14 REV. 12, JANUARY 2007 Regulatory Guide 1.14, Rev. 1 (August 1975)

Reactor Coolant Pump Flywheel Integrity Project Position - Not applicable to BWRs.

Regulatory Guide 1.15, Rev. 1 (December, 1972)

Testing of Reinforcing Bars for Category I Structures Project Position - Comply. USAR Section - Appendix B Regulatory Guide 1.17, Rev. 1 (June 1973)

Protection of Nuclear Power Plants Against Industrial Sabotage Project Position - Clinton Power Station complies with Regulatory Guide 1.17 with the following clarifications: 1)

Reference:

Paragraph C.1.b of the Regulatory Guide and GSA Interim Federal Specifications W-A-00450 B (GSA-FSS) dated February 6, 1973 - See Section 5.2.2.1 of the Physical Security Plan for commitments in alarm systems. 2)

Reference:

ANSI-N18.17-1973 - The Clinton Power Station Physical Security Plan sets forth the principles, policies, and general requirements for security at CPS. Because of its sensitive contents, any deviations are elaborated in the plan and thus are not provided in this section. It is the intent of the plan to meet the more recent requirements delineated in 10 CFR 73.55 and, due to considerable differences in scope between ANSI-N18.17-1973 and 10 CFR 73.55, the Physical Security Plan incorporates both documents to establish an effective

security policy to protect CPS against attempts of radiological sabotage. USAR Section - 13.6 CPS/USAR CHAPTER 01 1.8-15 REV. 11, JANUARY 2005 Regulatory Guide 1.18. Rev.1 (December, 1972)

Structural Acceptance Test for Conc rete Primary Reactor Containments Project Position - The Project complies with NRC Regulatory Guide 1.18 with the following exceptions: (1)

Reference:

Paragraph C.1 - Due to the low design pressure of 15 psig for the containment, raising the internal pressure of the containment in three approximately equal pressure increments is considered adequate for the structural acceptance test of the containment. (2)

Reference:

Paragraph C.3.- Tangential deflections around the equipment hatch which is the largest opening in the containment are insignificant. Therefore, only the radial deflections will be measured around the equipment hatch. USAR Subsection - 3.8.1.7.1 Regulatory Guide 1.19, Rev. 1 (August 11, 1972)

Nondestructive Examination of Primary Containment Liner Welds Project Position - The Project complies with NRC Regulatory Guide 1.19 with the following exception: (1)

Reference:

Paragraph C.7.b - Examinations by ultrasonic, magnetic particle and liquid penetrant methods were considered acceptable provided the examinations met the acceptance standards of NE 5330, NE 5340 and NE 5350 respectively of Section III of the 1971 ASME Code (Summer '73 Addenda). USAR Section - Appendix B.

Regulatory Guide 1.20, Rev. 2 (May 1976)

Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing Project Position - Comply USAR Subsections - 3.9.2.4, 14.2.12.1.37 CPS/USAR CHAPTER 01 1.8-16 REV. 11, JANUARY 2005 Regulatory Guide 1.21, Rev. 1 (June 1974)

Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants Project Position - The Project complies with NRC Regulatory Guide 1.21 with the following exceptions: (1)

Reference:

Paragraph C.6 - Turbulent flow in line 0WE32AA-4 causes thorough mixing of the fluid. A representative sample will still be obtained by sampling from the pipe because of this mixing. (2)

Reference:

Paragraph C.1 - A summary report of the meteorological measurements taken during each calendar quarter of the January through June period shall be provided as joint frequency distributions of wind direction and wind speed by atmospheric stability class in the July to December report as part of an annual summary. CPS will retain the January through June summary on site in a file that shall be provided to the NRC upon request. (3)

Reference:

Paragraph C.2 - CPS will make measurement of effluent volume, rates of release, and specific radionuclides for gaseous releases from the Station Heating, Ventilating, and Air Conditioning (HVAC) and Standby Gas Treatment System (SGTS) stacks. (4)

Reference:

Appendix A, Paragraph A.1.a. - CPS will perform gaseous grab samples on the HVAC stack and analyze for principal gamma emitters and tritium weekly and following reactor shutdown, startup, or a thermal power change exceeding 15% of rated thermal power within a one-hour period. CPS will perform gaseous grab samples on the SGTS stack upon the initiation of gaseous releases via the SGTS stack. (5)

Reference:

Appendix A, Paragraph B.1.b - A quarterly sample composited from proportional aliquots from each liquid effluent batch release during the quarter will be analyzed for tritium and gross alpha radioactivity. (6)

Reference:

Appendix B, Paragraph A.2 - The effluent concentrations of 10CFR20 are not utilized directly for limiting gaseous effluents. The CPS Offsite Dose Calculation Manual establishes requirements to limit the release rate of effluents. Discharges of gaseous radioactive material will not result in annual average exposure concentrations greater than limits for a member of the public in an unrestricted area (inside or outside the site boundaries). (7)

Reference:

Appendix B, Paragraph E.2, 5 and 6 - Doses at CPS are calculated in accordance with the NUREG-0133 maximum exposed individual concept. Dose due to the release of radioactive material in waterborne effluents is calculated for the water related pathways as specified in the CPS Offsite Dose Calculation Manual. Beta and gamma air dose due to the release of noble gas in gaseous effluents is calculated at the CPS site boundary in each of the 16 geographical directions surrounding CPS. Dose due to the release of radioactive iodines and particulates in gaseous effluents is calculated at the critical receptor location in each of the 16 geographical sectors surrounding CPS to a distance of CPS/USAR CHAPTER 01 1.8-17 REV. 11, JANUARY 2005 Regulatory Guide 1.21, Rev. 1 (June 1974) 5 miles. Dose summaries based on these calculations are provided in the

Radioactive Effluent Release Report. (8)

Reference:

Appendix B, Paragraph A.3 - The CPS Offsite Dose Calculation Manual limits the dose rates due to the release of fission and activation gases to less than or equal to 500 mrem per year to the total body and less than or equal to 3000 mrem per year to the skin. Release rates for fission and activation gases in the gaseous effluents are not determined directly from the average energy

)( of the radionuclide mixture in the effluent. Therefore, the

)( value for the gamma and beta energies per disintegration is not reported in Radioactive Effluent Release Reports. USAR Section/Subsections - 7.1.2.6.4, 7.6.1.2.4, 7.6.1.2.5, 7.6.1.2.6, 7.6.2.2.4, 7.6.2.2.5, 7.7.2, 9.3.2, 11.5 Regulatory Guide 1.22, Rev. 0 (February 1972)

Periodic Testing of Protection System Actuation Functions Project Position - Comply USAR Subsections - 7.1.2.6.5, 7.2, 7.3, 7.3.2.2.2.1.2, 7.4, 7.6, 8.1.6.1.3.

Regulatory Guide 1.23, Rev. 1 (Proposed)

Onsite Meteorological Program Project Position - Meets the requirements of ANS 2.5-1984 proposed as Regulatory Guide 1.23, Revision 1, with the following exceptions: (1) accuracy of the dewpoint temperature; (2) precipitation is not recorded on the digital portion of the data acquisition system; (3) digital accuracies.

USAR Subsection - 2.3.3 CPS/USAR CHAPTER 01 1.8-18 REV. 12, JANUARY 2007 Regulatory Guide 1.24, Rev. 0 (March 1972)

Assumptions Used for Evaluating the Potential Radiological Consequences of Pressurized Water Reactor Radioactive Gas Storage Tank Failure Project Position - Not applicable to BWRs. USAR Subsection - N/A Regulatory Guide 1.25, Rev. 0 (March 1972)

Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors Project Position - Partially Comply; for Alternative Source Terms, Regulatory Guide 1.183 is also utilized. USAR Section - 15.7.4 CPS/USAR CHAPTER 01 1.8-19 REV. 12, JANUARY 2007 Regulatory Guide 1.26, Rev. 3 (February 1976)

Quality Group Classifications and Standards for Water-, Steam-, and Radioactive- Waste-Containing Components of Nuclear Power Plants Project Position - The Project complies with NRC Regulatory Guide 1.26 with the following clarifications: (1)

Reference:

Paragraph C.2.a - The component cooling water system (CCW), which cools the fuel pool cooling and cleanup system (FPC&C) heat exchangers during normal operation is not classified as Quality Group C. However, that portion of the CCW system which includes the FPC&C heat exchanger inlet and outlet piping and valves which allow isolation from the remainder of the system is designed to Quality Group C standards. Essential service water is used to cool the FPC&C heat exchangers through this piping following a LOCA or LOEP. (2)

Reference:

Paragraph C.2.b - The reactor recirculation pumps are not considered essential and therefore are not provided with Quality Group C cooling water during normal operation and following a LOCA. The recirculation pump seals and motor bearings are, however, provided with cooling water following a loss of offsite electrical power in order to minimize any damage during coastdown. (3)

Reference:

Paragraph C.2.b - The cooling coils fitted in the thrust bearing housing of the Division 1 and Division 2 shutdown service water pump motors are provided with Quality Group C cooling water during normal operation and following a LOCA. The motor and its cooling coil are not assigned to quality group classification. For further description of classification see USAR Table 3.2-

1. (4)

Reference:

Paragraph C.2.b - The portion of Shutdown Service Water System (SSWS) piping and components associated with Standby Gas Treatment System (SGTS) Exhaust HI-Range Radiation Monitor Cooler are not classified as Quality Group C. However that portion is designed to Quality Group C standards.

USAR Section - 3.2, 5.4.7.1.1.6.1, 11.2.1.3, 9.2.1.2.1.1, 9.2.1.2.3

CPS/USAR CHAPTER 01 1.8-20 REV. 11, JANUARY 2005 Regulatory Guide 1.27, Rev. 2 (January, 1976)

Ultimate Heat Sink for Nuclear Power Plants Project Position - The Project complies with NRC Regulatory Guide 1.27 with the following exceptions: (1)

Reference:

Paragraph C.1.- The time period for the analysis of the ultimate heat sink was based upon the worst conditions for heat transfer from Clinton Lake.

These worst conditions were determined by a computer analysis of the lake which predicted transient lake temperature for the years 1949 through 1971.

Although two separate periods were not chosen for analysis, the intent of the Regulatory Guide to assure the adequacy of the ultimate heat sink during periods of high evaporative conditions and during period of unfavorable ambient conditions is met. (2)

Reference:

Paragraph C.2.C - The failure of the ultimate heat sink dam is not considered credible. USAR Subsections - 2.4.11.6, 9.2.5; TS3.7.1, 3.7.2 and 5.5.12 Regulatory Guide 1.28, Rev. 3 (August 1985)

Quality Assurance Program Requirements (Design and Construction)

Project Position - Comply with the following clarification:

The site QA programs, as committed in the PSAR were in compliance with ANSI N45.2(1971) as endorsed by Regulatory Guide 1.28 Rev. 0 dated June 7, 1972. Later revisions of ANSI N45.2 and the associated Regulatory Guide are incorporated in site QA programs. USAR Section - 17.1 CPS/USAR CHAPTER 01 1.8-21 REV. 11, JANUARY 2005 Regulatory Guide 1.29 Rev. 3 (September 1978)

Seismic Design Classification Project Position - The Project complies with NRC Regulatory Guide 1.29 with the following clarifications: 1)

Reference:

Paragraph C.1.e - The main steam system from the outermost containment isolation valve up to and including the shut-off valve in the Seismic Category I Auxiliary Building is classified and designed as Seismic Category I.

The portion of the system from the main steam shut-off valve up to the boundary of the Auxiliary Building is not classified as Seismic Category I, but is designed seismically. Seismic interface restraints are provided for each line near the Auxiliary/Turbine Building boundary. The piping from the Auxiliary/Turbine Building boundary up to and including the turbine stop valve is not classified or designed as Seismic Category I. The piping from the shut-off valve to the turbine stop valve is classified as Quality Group D in accordance with Regulatory Guide 1.26. 2)

Reference:

Paragraph C.1.h - The reactor recirculation pumps are not considered essential and therefore are not provided with Seismic Category I cooling water during normal operation and following a LOCA. The recirculation pump seals and motor bearings are, however, provided with cooling water following a loss of offsite electrical power in order to minimize any damage during coastdown. 3)

Reference:

Paragraph C.3 - The seismic design requirements for piping and supports beyond the defined Seismic Category I boundaries are described in Subsection 3.7.3.13. USAR Section/Subsection - 3.2, 3.7.3, 5.4.7.1.1.6.1, 7.1.2.6.6, 7.3.2.3.2.1.3, 7.3.2.20.2.1.2, 7.6.2.5.5 CPS/USAR CHAPTER 01 1.8-22 REV. 11, JANUARY 2005 Regulatory Guide 1.30, Rev. 0 (August 1972)

Quality Assurance Requirements for the Installation, Inspection and Testing of Instrumentation and Electrical Equipment Project Position - The project complies with the requirements of Regulatory Guide 1.30 with the following exception:

Reference:

ANSI N45.2.4-1972/IEEE 336-1971, Section 6.2.1 It is specified that for installed equipment, "items requiring calibration shall be tagged or labeled on completion indicating date of calibration and identity of person that performed the calibration." Complying with an obligation of this nature would be very restrictive and costly. It should clearly be noted that compliance is possible, but is neither necessary nor desirable since instrument ID numbers permit traceability to calibration records. The status of calibration of all permanent ly installed equipment covered by the Clinton Power Station "CALIBRATION PROGRAM" will be easily attained due to the traceability of records to that item. USAR Subsections - 6.5.1.5, 7.1.2.6.7, 7.3.2.2.2.1.4, 7.3.2.20.2.1.3, 7.4.2.1.2.1.5, 7.4.2.2.2.1.5, 7.6.2.4.2.1, 7.6.2.7.2, 7.6.2.8.2, 8.1.6.1.5, 8.1.6.2.6.

Regulatory Guide 1.31, Rev. 3 (April 1978)

Control of Ferrite Content in Stainless Steel Weld Metal Project Position for Balance of Plant Systems Scope of Supply - Comply Project Position for Nuclear Steam Supply Systems Scope of Supply - Compliance evaluation was based upon a comparison of the work performed for this project against the requirements of Revision 2 of Regulatory Guide 1.31. Compliance with this regulatory guide is based upon an extensive test program which demonstrates that controlling weld filler metal ferrite at 5% minimum produces production welds which meet the regulatory guide requirements. Reactor internals were fabricated prior to the issuance of Revision 3, however, ferrite measurements were made in accordance with the requirements of the ASME code in effect at that time. USAR Subsections - 4.5.1.2, 4.5.2.4, 5.2.3.4, 6.1.1, 10.3.6 CPS/USAR CHAPTER 01 1.8-23 REV. 14, JANUARY 2011 Regulatory Guide 1.32, Rev. 2 (February, 1977)

Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants Project Position

- Comply, with the following exceptions: 1)

Reference:

Paragraph C.1.c - Battery performance discharge tests are performed as required in TS 3.8.4, TS 3.8.5, and TS 3.8.6. Position C.1.e - with the clarifications and exceptions noted in subsection 8.1.6.1.6. 2)

Reference:

Paragraph C.1.c - IEEE Standard 450-1995 revision is used in lieu of earlier revisions. USAR Subsections - 7.1.2.6.8, 7.3.2.1.2.1.6, 7.3.2.20.2.1.4, 7.4.2.1.2.1.6, 8.1.6.1.6, 8.3.1.1.2.1, 8.3.1.2.2, 8.3.2.1.2.1, 8.3.2.2.2.2, TS 3.8.4, 3.8.5, 3.8.6.

Regulatory Guide 1.33, Rev. 2 (February 1978)

Quality Assurance Program Requirements (Operation)

Project Position - The Project complies with this guide with the following clarifications: 1)

Reference:

Paragraph C.2 - CPS compliance positions to ANSI Standards included and referenced an ANSI N18.7-1976 are addressed under the appropriate Regulatory Guide listed in this section. 2)

Reference:

Section 5.2.13.2 of ANSI N18.7-1976, Control of Purchased Material, Equipment and Services, 4th paragraph, 1st sentence - The phrase "...prior to installation or use of such items." is considered to be too restrictive to the efficient utilization of the plant. The following alternate course of action will provide the needed controls and ensure that items are not "released for operation" until the documentary evidence is available at CPS or an engineering evaluation has been performed, reviewed and accepted. Inservice items that are found to be nonconforming shall be reviewed to determine equipment operability as defined by the Technical Specifications. For items that represent significant conditions adverse to quality or safety, or require a repair or use-as-is disposition, an engineering evaluation shall be performed. The engineering evaluation shall provide support for the initial operability decision and provide the correction or resolution for the identified nonconformance. These items shall be controlled in accordance with approved procedures. Installed items not inservice that are nonconforming or become nonconforming as a result of maintenance shall be corrected or resolved prior to operational reliance. These items shall be controlled in accordance with approved

procedures.

CPS/USAR Regulatory Guide 1.33, Rev. 2 (February 1978) (Cont'd) CHAPTER 01 1.8-24 REV. 11, JANUARY 2005 A nonconforming item may be conditionally released for fabrication, installation or testing following an engineering evaluation to determine if such a conditional release is not detrimental to other components or systems. Conditional release items are controlled in accordance with approved procedures. The nonconformance for the conditionally released item shall be corrected or resolved prior to operational reliance. 3)

Reference:

Section 6 of ANSI N18.7-1976, References Subsequent revisions to the American National Standards referred to in ANSI N18.7-1976 will be evaluated to consider the necessity for incorporation of the revision into the CPS Operational Quality Assurance Program. 4)

Reference:

Paragraph C.1 3rd sentence - The following will be used in place of Appendix A to determine if a procedure is Safety-Related. A procedure shall be considered Safety-Related if the procedure operates, performs maintenance on, installs, modifies, or maintains the integrity of the pressure boundary for any system, components, or structure with a Safety Classification listed in Table 3.2-1 of the USAR and the Administrative Procedures listed in Appendix A, Item 1 of this Regulatory Guide. 5)

Reference:

Paragraphs C.5.i and C.5.j - The formats described in Section 13.5.2.1.3 shall be used in place of the format described in Section 5.3.9 of ANSI

N18.7-1976. 6)

Reference:

Section 5.2.6 of ANSI N18.7-1976, Equipment Control, 5th paragraph - Startup complies with the administrative controls for temporary modifications such as temporary bypass lines, electrical jumpers, lifted electrical leads and temporary trip point settings except that during the Checkout and Initial Operation (C&IO) Test Phase, temporary modifications shall not require independent verification. Documented verification of restoration shall be provided, but may be provided by the individual restoring the modification. 7) With regard to Section 5.2.15 of ANSI N18.7-1976 titled Review Approval and Control of Procedure: Programmatic controls for periodic reviews of procedures will consist of four key elements for the periodic review process:

• At least every two years Nuclear Oversight shall audit a representative sample of routine plant procedures that are used more frequently than every two years,

• All applicable plant procedures shall be reviewed following an unusual incident or unexpected transient, operator error, and following a modification,

• Routine plant procedures that have been used at least biennially receive scrutiny by individuals knowledgeable in the procedures, and are updated as necessary to ensure adequacy during suitable controlled activities, and, CPS/USAR Regulatory Guide 1.33, Rev. 2 (February 1978) (Cont'd) CHAPTER 01 1.8-25 REV. 11, JANUARY 2005

• Routine plant procedures that have not been used for two years will be reviewed before use to determine if changes are necesary or desirable. The third element is an acceptable method to review procedures because the procedure is tested through actual use. This satisfies the intent of the review criteria in ANSI 18.7-1976, Section 5.2.15. 8)

Reference:

Paragraphs C.4.a, C.4.b, C.4.c - The following audit frequencies will be followed: a. Audits of the results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or method of operation that affect nuclear safety - 24 months. b. Audits of the conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions - 24

months. c. Audits of the performance, training and qualifications of the unit staff - 24 months. 9)

Reference:

ANSI N18.7-1976, Section 5.2.2, Procedure Adherence, sentence 4, prescribes that one of the approvers of a temporary change to a procedure shall be the supervisor in charge of the shift and hold a senior operator license on the unit affected. The station was originally licensed with SAR section 13.5 requiring one of the approvers to hold a senior operators license on the unit affected, but did not specify that individual be the supervisor in charge of the shift. A CPS administrative procedure specifies that one of the unit management staff members approving temporary procedure changes shall be from the on-duty shift and hold an SRO license.

USAR subsection - 12.5.2, 13, 17.2, TS 5.4.1a Regulatory Guide 1.34, Rev. 0 (December 1972)

Control of Electroslag Weld Properties Project Position - Not applicable, electroslag welding not utilized. USAR Subsections - 5.2.3.3.2.2 Regulatory Guide 1.35, Rev. 3 (April, 1979)

Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures Project Position - Not applicable to Clinton Power Station.

CPS/USAR CHAPTER 01 1.8-26 REV. 11, JANUARY 2005 Regulatory Guide 1.36, Rev. 0 (February, 1973)

Nonmetallic Thermal Insulation For Austenitic Stainless Steel Project Position - Comply USAR Subsections - 5.2.3.2.4, 6.1.1, 10.3.6 Regulatory Guide 1.37, Rev. 0 (March 1973)

Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants Project Position - Comply USAR Subsections - 6.1.1, 10.3.6, 14.2.7 CPS/USAR CHAPTER 01 1.8-27 REV. 11, JANUARY 2005 Regulatory Guide 1.40, Rev. 0 (March 1973)

Qualification Tests of Continuous - Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants Project Position - Comply USAR Section - 7.1.2.6.9, 8.1 Regulatory Guide 1.41, Rev. 0 (March 1973)

Preoperational Testing of Redundant Onsite Electrical Power Systems to Verify Proper Load Group Assignments Project Position - Comply USAR Chapter - 8.3.1.2.1, 14 Regulatory Guide 1.42 has been withdrawn Regulatory Guide 1.43, Rev. 0 (May.1973)

Control of Stainless Steel Cladding of Low-Alloy Steel Components Project Position - Regulatory Guide 1.43 prescribes qualification and production cladding controls for ASME SA 508-2 material made to COARSE GRAIN practice. This material is not used for any of the safety clas s components. ASME SA 508-2 composition material employed on the reactor pressure vessel for this plant is produced to FINE GRAIN practice. Therefore, this Regulatory Guide is not applicable to the components in this plant. USAR Subsection - 5.2.3.4, 5.3.1.4.1.3, 6.1.1 CPS/USAR CHAPTER 01 1.8-28 REV. 11, JANUARY 2005 Regulatory Guide 1.44, Rev. 0 (May, 1973)

Control of the Use of Sensitized Stainless Steel Project Position for Balance of Plant Systems Scope of Supply - The requirements of Regulatory Guide 1.44 are met except as follows: 1.

Reference:

Paragraph C.3. Testing is not performed. Material is procured in the solution annealed conditon. 2.

Reference:

Paragraph C.6. All welding performed on austenitic stainless steel is with low heat input welding processes. Materials used are in the solution annealed condition and the following additional safeguards are taken: a. The preheat and interpass temperature used during the welding of austenitic stainless steel is kept to 350 degrees F maximum. b. Postweld heat treatment in the range of 800 degrees to 1500 degrees F is strictly forbidden. Solution annealing heat treatment, after welding, although not required, is permitted. Because severe sensitization is avoided by these safeguards, testing to determine susceptibility to intergranular attack is not performed. Project Position for Nuclear Steam Supply System Scope of Supply - Complies with the intent of the Regulatory Guide 1.44 USAR Subsections - 5.2.3.4, 6.1.1, 10.3.6.

CPS/USAR CHAPTER 01 1.8-29 REV. 12, JANUARY 2007 Regulatory Guide 1.45, Rev. 0 (May, 1973)

Reactor Coolant Pressure Boundary Leakage Detection Systems Project Position - The requirements of the NRC Regulatory Guide 1.45 are met with the following exceptions for those portions of the station under the balance of plant systems scope of design: 1.

Reference:

Paragraph C.5 - The sensitivity and response time of airborne particulate and gaseous radioactivity monitors is not adequate to detect a leakage rate of 1 GPM in less than one hour. The correlation between flow rate and radioactivity is not valid due to various complex factors discussed in Section 5.2.5.2.2. The monitor will not always alarm for 1 GPM in one hour and, therefore, is considered as qualitative indication of the presence of abnormal leakage. Similarly, because the drywell floor drain sump flow monitoring system calculates the unidentified leakage rate based on drywell floor drain sump pump discharge flow, this system is not sensitive enough to detect a leakage rate of 1 GPM in less than one hour over the entire range of potential leakages. This

system is, however, capable of promptly detecting leakage rates and leakage rate increases prior to exceeding Technical Specification operating limits. CPS will follow the guidelines of ANSI/ISA S67.03, Standard for Light Water Reactor Coolant Pressure Boundary Leak Detection, October 3, 1982. 2.

Reference:

Paragraph C.6 - The sump flow monitoring instrumentation which is located at the sumps is seismically qualified to OBE. The calculation devices and instrumentation outside of the drywell have not been seismically qualified. Should these devices fail during a seismic event, they are readily accessible for maintenance and/or replacement to reestablish the functionality of the monitoring

equipment. 3.

Reference:

Paragraph C.3 - Compliance is met through the implementation of NUREG 1434, Standard Technical Specifications for BWR-6 plants, LCO 3.4.7, Reactor Coolant Systems Leakage Detection Instrumentation, which was approved by the NRC in License Amendment 95. USAR Subsections - 5.2.5, 7.3, 7.4, 7.6, 7.6.1.4, 7.6.2.4, TS 3.4.7 Regulatory Guide 1.46 has been withdrawn.

CPS/USAR CHAPTER 01 1.8-30 REV. 11, JANUARY 2005 Regulatory Guide 1.47, Rev. 0 (May 1973)

Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems Project Position - Comply USAR Subsections - 7.1.2.6.11, 7.2, 7.3, 7.3.2.20.2.1.5, 7.4, 7.6, 8.1.6.1.9, 8.3.2.2.2.3, App. D Regulatory Guide 1.48, Rev. 0 (May 1973)

Design Limits and Load Combinations for Seismic Category 1 Fluid System Components Project Position - Comply with stress limits for Active Components with the following exception: The operability requirements for all active components will be assured by performing a detailed deformation analysis and/or by performing a seismic test. Therefore, the allowable stress limits that shall be used for each category shall be in conformance with the applicable ASME Codes. USAR Section - 3.9 Regulatory Guide 1.49, Rev. 1 (December 1973)

Power Levels of Nuclear Power Plants Project Position - Comply USAR Sections - 15.0, 15.1 CPS/USAR CHAPTER 01 1.8-31 REV. 11, JANUARY 2005 Regulatory Guide 1.50, Rev. 0 (May 1973)

Control of Preheat Temperature for Welding of Low-Alloy Steel Project Position for Balance of Plant Systems Scope of Supply - Low-alloy steels were not used on ASME Class 1, 2 and 3 piping systems during plant construction, therefore, control of preheat temperature for welding as required by Regulatory Guide 1.50 was not applicable to these systems at that time. Monitoring of plant operation has revealed certain sections of piping to be susceptible to Flow Accelerated Corrossion (FAC). Low-alloy steels, such as 21/4 Cr - 1 Mo, may be used as repair/replacement materials is these piping sections. Where low-alloy steel is used the requirements of Regulatory Guide 1.50 for control of welding preheat temperature will be complied with. Project Position for Nuclear Steam Supply Systems Scope of Supply

- During plant construction the use of low-alloy steel was restricted to the reactor pressure vessel. For fabrication of the reactor pressure vessel welding preheat control complied with Regulatory Guide 1.50. Low-alloy steels were not used on the remainder of NSSS systems during plant construction, therefore, control of preheat temperature for welding as required by Regulatory Guide 1.50 was not applicable to these systems at that time. Monitoring of plant operation has revealed certain sections of piping to be susceptible to Flow Accelerated Corrosion (FAC). Low-alloy steels, such as 21/4 Cr - 1 Mo, may be used as repair/replacement materials is these piping sections. Where low-alloy steel is used the requirements of Regulatory Guide 1.50 for control of welding preheat temperature will be complied with. USAR Subsections - 5.2.3.3.2.1, 5.3.1.4.1.5, 10.3.6 Regulatory Guide 1.51 has been withdrawn Regulatory Guide 1.52, Rev. 2 (March 1978)

Design, Testing and Maintenance Criteria for Post Accident Engineered - Safety - Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants Project Position - As required by Regulatory Guide 1.70, Revision 3, a detailed discussion of the extent of compliance to the requirements of Regulatory Guide 1.52 is provided in USAR Table 6.5-3. In addition, the CPS "ANSI N509/510 Variance Report",

identifies minor deviations to this Regulatory Guide referenced in the ANSI standards.

This Variance Report is a controlled document, maintained by the Nuclear Station Engineering Department. USAR Section - 6.5, 9.5.1.2.2.3, 12.3.3.1, E4.0.D.4.d, Table 7.1-3, Table 14.2-1

CPS/USAR CHAPTER 01 1.8-32 REV. 11, JANUARY 2005 Regulatory Guide 1.53, Rev. 0 (June 1973)

Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems Project Position - Comply USAR Subsections - 7.1.2.6.12, 7.3.2.3.2.1.5, 7.3.2.20.2.1.6, 7.6.2.5.5, 8.1.6.1.10 Regulatory Guide 1.54, Rev. 0 (June 1973)

Quality Assurance Requirements for Protective Coatings Applied to Water Cooled Nuclear Power Plants Project Position - Comply USAR Sections - 6.1, 6.2, 9.1.4.2.5.7, 17.1, 17.2 Regulatory Guide 1.55, Rev. 0 (June 1973)

Concrete Placement in Category I Structures Project Position - Comply USAR References - Appendix B and Section 17.1 Regulatory Guide 1.56, Rev. 1 (July 1978)

Maintenance of Water Purity in Boiling Water Reactors Project Position - Comply with the following exceptions: 1.

Reference:

Paragraph C.4d - The resin samples will be taken every 5th removal from service for resin cleaning. Since most units may be removed from service one or more times for resin cleaning with remaining capacity available, they will be returned to service without regeneration. To count such removals from service may mean determining remaining capacity at a time the bed would not normally be considered at "minimum residual capacity". The resin beds will be replaced when the calculated resin capacity approaches 50% of initial capacity. 2. Condensate polishing resin beds may be used beyond 50% of initial capacity while the reactor is shutdown, circ water is isolated and feedwater is not being supplied to the vessel via the condensate polishers. USAR Subsections - 5.2.3.2, 10.4.6, 12.3 CPS/USAR CHAPTER 01 1.8-33 REV. 11, JANUARY 2005 Regulatory Guide 1.57, Rev. 0 (June 1973)

Design Limits and Loading Combinations for Metal Primary Reactor Containment System Component Project Position - The design complies with Regulatory Guide 1.57 with the following clarification: For Class MC penetration assemblies with respect to Regulatory Guide Position C.1.d, the design load combination for faulted loads are: a. Maximum operating pressures and temperatures, plus loads due to pipe rupture and jet impingement where applicable. b. Process pipe maximum operating pressure applied in the annulus between the process pipe and the penetration sleeve for MC penetration assemblies only. USAR Sections - 3.8, 3.9 Regulatory Guide 1.59, Rev. 2 (August, 1977)

Design Basis Floods for Nuclear Power Plants Project Position - Comply USAR Subsections - 2.4.2.3, 2.4.3, 2.4.8 Regulatory Guide 1.60, Rev. 1 (December, 1973)

Design Response Spectra for Seismic Design of Nuclear Power Plants Project Position - Comply. USAR Subsections - 2.5.2, 3.7.1 Regulatory Guide 1.61, Rev. 0 (October 1973)

Damping Values for Seismic Design of Nuclear Power Plants Project Position - Comply USAR Subsections - 3.7.1, 3.7.2, 3.7.3 CPS/USAR CHAPTER 01 1.8-34 REV. 11, JANUARY 2005 Regulatory Guide 1.62, Rev. 0 (October 1973)

Manual Initiation of Protective Actions Project Position - Comply USAR Subsections - 7.1.2.6.14, 8.1.6.1.11.

Regulatory Guide 1.63, Rev. 2 (July, 1978)

Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants Project Position - Comply. USAR Subsections - 8.1.6.1.12.

Regulatory Guide 1.65, Rev. 0 (October 1973)

Materials and Inspections for Reactor Vessel Closure Studs Project Position - Comply USAR Subsection - 5.3.1.7 Regulatory Guide 1.66 has been withdrawn.

CPS/USAR CHAPTER 01 1.8-35 REV. 11, JANUARY 2005 Regulatory Guide 1.67, Rev. 0 (October 1973)

Installation of Overpressure Protective Devices Project Position - The main steam line safety/relief valves relieve to closed discharge systems. The guidelines delineated in Regulatory Guide 1.67 are not applicable to that situation. The installation of safety/relief valves in other ASME Class 1 and 2 systems complies with the regulatory guide. USAR Subsections - 3.9.3.3.2, 5.2.2 CPS/USAR CHAPTER 01 1.8-36 REV. 11, JANUARY 2005 Regulatory Guide 1.68, Rev. 2 (August 1978)

Initial Test Programs for Water-Cooled Nuclear Power Plants Project Position - The Project complies with the following exceptions. 1.

Reference:

Appendix A, Paragraph 1H(10) - There is no practical way of lowering the lake level for testing of the ultimate heat sink. Testing for NPSH to ensure vortexing does not occur will be performed by lowering the water level in the pump pit. Testing will be performed at simulated normal lake level and minimum design ultimate heat sink level. 2.

Reference:

Appendix A, Paragraphs 1k(2) and (3) - There will be no "preoperational test" of personnel radiation monitoring instruments or laboratory radiation measuring equipment. Equipment of this type will be calibrated to assure proper operation. 3.

Reference:

Appendix A, Paragraph 5k - This test will not be performed during power operation since this would induce an unnecessary thermal transient cycle to the nozzles and spargers. The high pressure coolant injection systems will be preoperationally tested to verify starting, flow rates and head loss. 4.

Reference:

Appendix A, Paragraph 5L - RHR System testing in Steam Condensing Mode was not performed during the Startup Test Program. If this mode of operation is ever used it will be evaluated for testing requirements at

that time. 5.

Reference:

Appendix A, Paragraph 5hh - Reactor coolant flow control system calibration and performance in the Automatic Load Following mode was not performed during the Startup Test Program. CPS operational programs were revised to delete this feature of recirc. control. The feature has been subsequently removed. 6.

Reference:

Appendix A, Paragraph 5x - A test to determine the heat removal capacity was not performed for the ECCS Equipment Cooling HVAC System. Heat removal adequacy was demonstrated for compliance to this paragraph. USAR Chapter

- 14 Regulatory Guide 1.68.1, Rev. 1 (January 1977)

Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants Project Position - Comply USAR Section - 14.2 CPS/USAR CHAPTER 01 1.8-37 REV. 14, JANUARY 2011 Regulatory Guide 1.68.2, Rev. 1 (July 1978)

Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants Project Position - Comply USAR Section - 14.2 Regulatory Guide 1.69, Rev. 0 (December 1973)

Concrete Radiation Shields for Nuclear Power Plants Project Position - Comply USAR Subsection - 12.3.2 Regulatory Guide 1.70, Rev. 3 (November, 1978)

Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants - LWR Edition Project Position - Comply with the following clarifications and exception: 1. Clarification: The FSAR was written in accordance with the guidance of Regulatory Guide 1.70, Rev. 3. However, CPS will utilize Regulatory Guide 1.181 in conjunction with NEI 98-03, Guidelines for Updating Final Safety Analysis Reports, as guidance for maintaining the USAR in accordance with the

requirements of 10 CFR 50.71(e). 2. Clarification: CPS is a single unit nuclear facility, owned and operated by Exelon. As such, Exelon has a corporate based organization directly involved in the operation and in providing technical or operational support to CPS. The description of the organizations providing technical or operational support to CPS including the corporate and site organizations are described in the Exelon Generation Company Quality Assurance Topical Report. 3. Exception: Reference Section 1.7.1. Consistent with Regulatory Guide 1.70. Electrical and Instrumentation and Cont rol drawings were provided to the NRC during initial licensing of the facility. However, per the Questions/Responses enclosed with Generic Letter 81-06 (see Question/Response C.1), these drawings are not included in the USAR. USAR Subsection - 1.1.9.2, 4.1.2.1.1.(5), 6.2.8.13, 6.3.2.8, 7.1.1.1, 7.1.2.6.17, 7.6.2.4.2.1, 7.6.2.7.2.1, 7.6.2.8.2.2, 7.7.2.9.2, .7.2.10.2, 17.0, Appendix D II.D.1, Appendix D II.K.3.44.

CPS/USAR CHAPTER 01 1.8-38 REV. 11, JANUARY 2005 Regulatory Guide 1.71, Rev. 0 (December 1973)

Welder Qualification for Areas of Limited Accessibility Project Position - For Clinton Power Station construction, requirements or Regulatory Guide 1.71 are met with the following exceptions: 1.

Reference:

Paragraph C.1 - Performance of welders under simulated access conditions is not necessary to assure acceptable welds. There would be an excessive number of required qualifications. Accept ability of welds will be

determined by required examinations. 2.

Reference:

Paragraph C.2 - Requalification is not necessary under the conditions listed in the regulatory position. This would require an excessive number of requalifications. Acceptability of welds will be determined by required examination. For Clinton Power Station operation, welder qualification for areas of limited accessibility during the operation stage complies with the intent of Regulatory Guide 1.71 through the implementation of the following program: If clearance around the production joint of at least 12 inches (except for minor obstructions) cannot be obtained, and clearance is less than 12 inches where obstructions exist which may prevent the welder from gaining an advantageous position for welding, the weld will be evaluated for limited access considerations by the supervisor mechanical. In this case, the structure to be welded, including its actual access limits, may be simulated. USAR Subsections - 5.3.1.4, 10.3.6 Regulatory Guide 1.72, Rev. 2 (November, 1978)

Spray Pond Plastic Piping Project Position - Not applicable; Clinton Power Station does not have spray ponds.

Regulatory Guide 1.73, Rev. 0 (January 1974)

Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants Project Position - Comply USAR Section - 8.1 CPS/USAR CHAPTER 01 1.8-39 REV. 13, JANUARY 2009 Regulatory Guide 1.75, Rev. 2 (September, 1978)

Physical Independence of Electric Systems Project Position - The project complies with NRC Regulatory Guide 1.75 with the clarifications and exceptions noted in Subsections 7.1.2.6.19 and 8.1.6.1.14. USAR Subsections - 7.1.2.6.19, 7.3.2.20.2.1.7, 7.6.2.5.5, 8.1.6.1.14, 8.3.1, E4.0.D.3.c Regulatory Guide 1.76, Rev. 1 (March, 2007)

Design Basis Tornado and Tornado Missiles for Nuclear Power Plants Project Position - Comply. USAR Subsection - 3.3.2, 3.5.1.4

• Note: Reg 1.76, Rev. 1 is effective at CPS from September 2007.

Regulatory Guide 1.77, Rev. 0 (May, 1974)

Assumptions Used for Evaluating a Control Rod Ejection Accident for PWRs Project Position - Not applicable to BWRs.

Regulatory Guide 1.78, Rev. 0 (June, 1974)

Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release Project Position - Comply. USAR Sections/Subsection: 2.2, 6.4, 9.4.1, 9.5.8.3 CPS/USAR CHAPTER 01 1.8-40 REV. 11, JANUARY 2005 Regulatory Guide 1.79, Rev. 1 (September 1975)

Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors Project Position - Not applicable to BWRs.

Regulatory Guide 1.80, Rev. 0 (June 1974)

Preoperational Testing of Instrument Air System Project Position - The project complies with the requirements of Regulatory Guide 1.80 with the following exception:

Reference:

C.5 - The Instrument Air System will meet the Class 'B' cleanliness requirements as defined in ANSI N 45.2.1-1973, "Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants."

Particulate contamination criteria will be established for portions of the instrument air system that supply air to selected active safety-related components to ensure operational reliability.

Reference:

C.8 - A loss-of-instrument-air supply test will be performed on the branches of the system which serve sa fety-related equpment, specifically those which supply the Automatic Depressurization System (ADS) and Non-ADS Low

Low Set Safety Relief Valves (LLS-SRV) valve operators with operating air. Testing will be conducted on a valve-by-valve basis for those branches which do

not serve safety-related equipment.

Reference:

C.10 - A loss-of-instrument-air supply test as described in C.10 will not be performed in the Instrument Air Preoperational Test. The testing required by C.8 and C.9 in combination with the Checkout and Initial Operation Testing of each valve will satisfy the position of C.10. USAR Section/Subsection - 9.3.1.4, 14.2.12.1.46.

Regulatory Guide 1.81, Rev. 1 (January, 1975)

Shared Emergency and Shutdown Electrical Systems for Multi-Unit Nuclear Power Plants Project Position - Does not apply since Clinton is a single unit plant.

CPS/USAR CHAPTER 01 1.8-41 REV. 11, JANUARY 2005 Regulatory Guide 1.82, Rev. 2 (May 1996)

Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident Project Position - Comply. USAR Subsections - 6.2.2.2, 6.3.2.2 Regulatory Guide 1.83, Rev. 1 (July 1975)

Inservice Inspection of PWR Steam Generator Tubes Project Position - Not applicable to BWRs.

CPS/USAR CHAPTER 01 1.8-42 REV. 11, JANUARY 2005 Regulatory Guide 1.84 Design and Fabrication Code Case Acceptability ASME Section III Division I Project Position

1. CPS will apply all applicable approved code cases, not limited to those listed in the latest revision of Regulatory Guide 1.84, as found necessary and proper for the construction activities. a. N-237 - Hydrostatic Testing of Internal Piping,Section III, Division 1. Exemption from hydrostatic testing of the portion of ASME Class 2 containment spray piping system extending downstream from the containment penetration to the containment spray device Reference is made to NRC's letter Docket Nos. 50-461/462 dated August 18, 1980, with expressed comment that this code case requires the examination of the exempted piping in accordance with NB 5200 even though the piping is ASME Class 2. b. N-241 - Hydrostatic Testing of Piping,Section III, Division 1. Exemption from hydrostatic testing of safety relief valve discharge piping from the main steam, HPCS, LPCS, RHR, and RCIC systems to the suppression pool; including the submerged portions and terminations.

Reference is made to NRC's letter Docket Nos. 50-461/462 dated August 18, 1980, which clarified the code case applicability only to the submerged portions of the above-cited systems. c. N-240 - Hydrostatic Testing of Open Ended Piping,Section III, Division 1. Exemption from hydrostatic testing of:

(1) Shutdown service water discharge to the ultimate heat sink beyond the last isolation valve. (2) Various diffusers, drains, and return piping to the several fuel pools in the containment and fuel buildings. (3) Suppression pool makeup system piping connecting upper containment pool with suppression pool. (4) Diesel oil system pump suction piping from and return piping to the diesel generator fuel oil storage tanks and the fuel oil day

tanks. (5) Standby liquid control system pump suction piping from storage tank to first isolation valves.

CPS/USAR Regulatory Guide 1.84 (Cont'd) CHAPTER 01 1.8-43 REV. 11, JANUARY 2005 (6) Pump suction and return piping to the RCIC storage tank. (7) HPCS, LPCS, RHR, RCIC, and suppression pool cleanup pump suction piping from suppression pool to first isolation valve outside containment and return lines to the suppression pool. The approval to use Code Case N-240, "Hydrostatic Testing of Open Ended Piping,Section III, Division 1," was aut horized by NRC for construction of above cited components with no limitations by letter Docket Nos. 50-461/462 dated

February 25, 1980. d. N-341 - Certification of Level III NDE Examiner Section III, Divisions 1 and 2. The approval to use Code Case N-341 without limitations was granted by NRC letter Docket No. 50-461 dated April 11, 1983. e. N-315 - Repair of Bellows,Section III, Division 1.

Repair of bellows under ASME Section III, Division 1 for ten guard pipe expansion bellows at CPS. The approval for use of Code Case N-315 was authorized by NRC letter Docket No. 50-461 dated September 27, 1983 with the condition that CPS submit the description of the repair as well as justification for

repairing the bellows rather than replacing them. Following receipt of NRC's

approval for the repair, but prior to making the repair, CPS is to provide NRC the results of the qualification on the full-scale facsimile bellows, including the design requirements to assure that the repai r meets the requirements of the design specification. 2. Illinois Power Company requested and received approval for application of the following code cases: a. N-356 - Certification Period for Level III NDE Personnel Section XI, Divisions 1, 2, and 3.

The approval to use Code Case N-356 was authorized by NRC letter Docket No. 50-461 dated April 11, 1983 with no limitations. b. N-397 - Alternate Rules to the Spectral Broadening Procedures of N-1226.3 for Classes 1, 2, and 3,Section III, Division 1.

The approval to use Code Case N-397 was authorized by NRC letter Docket No. 50-461 dated April 5, 1985 with no additional requirements other than those specified in the code case. c. N-411 - Alternate Damping Values for Seismic Analysis of Piping Section III, Division 1, Class 1, 2 and 3 Construction. The approval of Code Case N-411 was conditionally granted by NRC letters dated July 19, 1985 and April 5, 1985. This code case will be used CPS/USAR Regulatory Guide 1.84 (Cont'd) CHAPTER 01 1.8-44 REV. 11, JANUARY 2005 for piping systems analyzed by response spectrum methods and not those using time-history analysis methods. This code case will be used, as necessary, for any future piping and equipment dynamic analysis or reanalysis. If, as a result of using the damping value in ASME Code Case N-411, piping supports are moved, modified or eliminated, any increased piping displacements due to the greater piping flexibility will be checked to assure that they can be accommodated and that there will be no adverse interaction with adjacent structures, components, or equipment. When the alternative damping values of this code case are used, they will be used in their entirety in a given analysis and not a mixture of Regulatory Guide 1.61 and Code Case N-411. This code case will be applied for both hydrodynamic loads as well as seismic loads; however, the damping values in Code Case N-411 are limited to a response frequency below 33 Hz. d. N-413 - Minimum Size of Fillet Welds for Linear Type Supports,Section III, Division 1, Subsection NF.

The approval to use Code Case N-413 was authorized by NRC letter Docket No. 50-461 dated April 30, 1985 with no limitations other than those stated in the text of the code case. 3. Acceptable code cases annulled by action of the ASME Council (or deleted in later revisions to this guide), but specified for procurement or other activities, shall remain valid. Commitment to meet a specific revision of this guide has little significance since the guide is revised as new code cases are issued by the ASME and approved by the NRC. USAR Subsection - 5.2.1.2 CPS/USAR CHAPTER 01 1.8-45 REV. 11, JANUARY 2005 Regulatory Guide 1.85 Materials Code Case Acceptability ASME Section III Division I Project Position

1. CPS will apply all applicable approved code cases, not limited to those listed in the latest revision of the Regulatory Guide 1.85, as found necessary and proper for the construction activities of the CPS. 2. Acceptable code cases annulled by action of the ASME Council (or deleted in later revisions to this guide), but specified for procurement, shall remain valid. Commitment to meet a specific revision of this guide has little significance since the guide is revised as new code cases are issued by the ASME and approved by the NRC. USAR Subsection - 5.2.1.2 Regulatory Guide 1.86, Rev. 0 (June 1974)

Termination of Operating Licenses for Nuclear Reactors Project Position - Comply.

Reference - License Application Regulatory Guide 1.87, Rev. 1 (June 1975)

Guidance for Construction of Class 1 Components in Elevated-Temperature Reactors Project Position - Not applicable to BWRs.

CPS/USAR CHAPTER 01 1.8-46 REV. 11, JANUARY 2005 Regulatory Guide 1.89, Rev. 0 (November 1974)

Qualification of Class 1E Equipment for Nuclear Power Plants Project Position - Comply USAR Section/Subsections - 3.11, 7.3.2.20.2.1.8, 7.6.2.5.5, 8.1 Regulatory Guide 1.90, Rev. 1 (August 1977)

Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons Project Position - Not applicable; Clinton Power Station has reinforced concrete containments.

Regulatory Guide 1.91, Rev. 1 (February, 1978)

Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants Project Position - Comply. USAR Subsection - 2.2.3 Regulatory Guide 1.92, Rev. 1 (February 1976)

Combining Modal Responses and Spatial Components in Seismic Response Analysis Project Position - The project complies with NRC Regulatory Guide 1.92 with the following clarification: In combining the modal response using double sum method, a corrected damping factor

'k is used for computing the damped frequency '

k of a system or a subsystem with closely spaced modes, while an uncorrected damping factor k is used per the regulatory guide. The justification for using

'k instead of k in the calculation of '

k is presented in the attached "Supplement to the Position on Regulatory Guide 1.92". USAR Section/Subsection - 3.7.2.7, 3.7.2.12, 3.7.3.7, 3.7.3.8 CPS/USAR CHAPTER 01 1.8-47 REV. 11, JANUARY 2005 Supplement to Position on Regulatory Guide 1.92 In the double sum method of modal combination, a modified damping factor, instead of an uncorrected value as in Equation (10) of Regulatory Guide 1.92, should be used for

the damped frequency to evaluate the correlation coefficient of the closely space modes. Equations (8), (9), (10) and (11) of the regulatory guide are based on a study by Rosenblueth and Elorduy (Reference 1) Referring to that paper, for a single-degree-of-freedom system governed by the equation of motion, ()()()()t x t q t q 2 t q 2 1 1 1=++ (1) The correction factor for damping can be expressed as (Equation (4) of Reference 1) 2/1 1 1)2/1 ()()(s Q Q+= (2) where s is the duration of a segment white noise excitation. E(Q) and E(Q

0) are the expected value of the dam ped and the undamped systems. The maximum response of a system to a transient disturbance of form x(t) = f(t) W(t) can be expressed (Equation (8) of Reference 1) dt q Q 2 0 2µ (3) The transfer function q(t) for the deformation of the system expressed by Equation (1) is (Equation 10.3 of Reference 2) y q(t)t'1sin t)11(exp'11= (4) where 2 1 1 1 1'= (5) When q is the pseudo-velocity of a single-degree system, the second member in Equation 4 gives 1/2 1 1. In order to adjust the percentage of damping to coincide with the expected response, Rosenblueth suggested the use of a modified damping factor (Equation (9) of Reference

1) s/2'1 1 1+= (6) in the system's natural mode of vibration.

CPS/USAR Supplement to Position on Regulatory Guide 1.92 (Cont'd) CHAPTER 01 1.8-48 REV. 11, JANUARY 2005 In other words, the uncoupled equation of motion of a multi-degree-of-freedom system should be adjusted as

)t (x a)t (q)t (q'2)t (q i i 2 i i i i i=++ (7) and the transfer function is given by (Equation 10.3 of Reference 2) t'sin)t'(exp'a i)t (q i i i i i= (8) where 2 i i i'1'= (9) Note here the modified damping factor

'i and not the uncorrected damping value i is used in Equation (9) for the damped frequency of the adjusted system. The final solution is then obtained based on the transfer function of Equation (8) but not Equation (4), as

)1 (/Q Q j i Q i Q 2 ij j i 2 i 2++= (10) where )/(j j i i j i ij+= (11) 2 i i i'1'= (12) Thus, a modified damping factor for the damped frequency should be used. Amin and Gungor (Reference 3) and Singh, Chu and Singh (Reference 4) also used the modified damping factor for the damped frequency in their computation of the correlation coefficent of closely spaced modes. For a lightly damped system and an earthquake duration of 10 seconds as in the Clinton design basis, the damped frequency based on a modified damping factor and the damped frequency based on a uncorrected damping factor are approximately the same.

However, on a theoretical basis, the modified damping factor should be used for the damped frequency in the evaluation of the correlation of the closely spaced mode response. For a 10 Hz system with 2% damping and 10 seconds earthquake duration, the damped frequency using modified and uncorrected damping factors is 9.9973 Hz and 9.9980 Hz respectively. Thus the use of either modified or uncorrected damping factor does not affect the results.

CPS/USAR Supplement to Position on Regulatory Guide 1.92 (Cont'd) CHAPTER 01 1.8-49 REV. 13, JANUARY 2009 References

1. E. Rosenblueth and J. Elorduy, "Response of Linear Systems to Certain Transient Disturbances" Proceedings, Fourth World Conference on Earthquake Engineering, Vol. 1 Santiago, Chile, 1969. 2. N. M. Newmark and E. Rosenblueth, "Fundamentals of Earthquake Engineering

", Prentice Hall Inc., 1971. 3. M. Amin and I. Gungor, "Random Vibration in Seismic Analysis - an Evaluation", ASCE National Structural Engineering Meeting , Baltimore, Maryland, April 1971. 4. A. K. Singh, S. L. Chu and S. Singh, "Influence of Closely Spaced Modes in Response Spectrum Method of Analysis", Proceedings of the Special Conference on Structural Design of Nuclear Plant Facilities, Chicago, IL, December 1973.

Regulatory Guide 1.93, Rev. 0 (December 1974)

Available Electric Power Sources Project Position - The project complies with the requirements of Regulatory Guide 1.93 with the following exception:

Reference:

Paragraphs C.1 and C.5 - The requirements provided in these paragraphs concerning the time limit for inoperability is not followed for: a) the Division 3 diesel generator and the Division 3 and 4 batteries. This exception is based on the facts that the only load on Division 3 is the high pressure core spray system (HPCS), and the Division 4 battery support for HPCS initiation. The CPS Technical Specifications require HPCS power source availability similar to the requirements contained in NRC Standard Technical Specification (NUREG 1434). b) the Division 1 and 2 diesel generators have allowed outage time (AOT) of 14 days rather than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and this time may be used for preventive maintenance. The basis for this change is risk-informed Techical Specification Amendment 141. USAR Subsection - 8.1.6.1.17; TS3.8.1, 3.8.4

CPS/USAR CHAPTER 01 1.8-50 REV. 14, JANUARY 2011 Regulatory Guide 1.96, Rev. 1 (June 1976)

Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants Project Position - Partially Comply. As a result of the re-analysis of the Loss-of-Coolant Accident (LOCA) using Alternative Source Term (AST) Methodology, it is no longer necessary to credit the Main Steam Isolation Valve Leakage Control System (MSIVLCS) for post-LOCA activity leakage mitigation. The system has been left in place as a passive system and is not required to perform any safety function. USAR Subsection - 5.4.7.1.1.6.1, 6.7.1.2 Regulatory Guide 1.97, Rev. 3 (May, 1983)

Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident Project Position - The project complies with the requirements of Regulatory Guide 1.97 with the clarifications and exceptions itemized in Table 7.1-13. In accordance with the SER to TS Amendment 164 Containment Hydrogen Monitoring is changed from a Category 1 variable to a Category 3 variable. USAR Tables - 7.1-13, 7.1-14, 7.5-1 USAR Sections - 7.1.2.6.23, 7.3.2.20.2.1.9, 7.6.2.12.4, 7.6.2.12.5, 7.7.1.26.3.6, 9.3.7.1.2, 12.5.2, Appendix D Regulatory Guide 1.98, Rev. 0 (March 1976)

Assumptions Used for Evaluating the Potential Radiological Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor Project Position - Comply USAR Subsection - 15.7.1 CPS/USAR CHAPTER 01 1.8-51 REV. 11, JANUARY 2005 Regulatory Guide 1.99, Rev. 2 (May 1988)

Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials Project Position - Comply USAR Subsections - 4.1.4.5, 4.3.2.8, 5.3.1.4, 5.3.1.5, 5.3.1.6, 5.3.2.1, and 5.3.2.2 Regulatory Guide 1.100, Rev. 1 (August 1977)

Seismic Qualification of Electric Equipment for Nuclear Power Plants Project Position - Comply USAR Section - 3.10, 6.2.5.2.3, 7.3.2.20.2.1.10, 7.6.2.5.5, 8.1.6.1.18, 8.3.1 Regulatory Guide 1.101, Rev. 2 (October 1981)

Emergency Planning for Nuclear Power Plants Project Position - Comply USAR Section - 13.3, E4.0.B.5 Regulatory Guide 1.102, Rev. 1 (September, 1976)

Flood Protection for Nuclear Power Plants Project Position - Comply. USAR Subsection - 3.4.1 Regulatory Guide 1.103, Rev. 1 (October 1976)

Post-Tensioned Prestressing Systems for Concrete Reactor Vessels and Containments Project Position - Since Clinton Power Station has reinforced concrete containments and steel reactor vessels, this regulatory guide does not apply.

CPS/USAR CHAPTER 01 1.8-52 REV. 11, JANUARY 2005 Regulatory Guide 1.104 has been withdrawn Regulatory Guide 1.105, Rev. 1 (November 1976)

Instrument Setpoints Project Position - The Project complies with NRC Regulatory Guide 1.105 with the following clarification:

Reference:

Paragraph C.5 - CPS shall comply with the regulatory positions established in this regulatory guide. However, some equipment setpoint adjustments are not mechanical and, therefore, do not have mechanical securing devices. The equivalent of mechanical securing devices are provided in the equipment design. USAR Section - 7.1.2.6.25, 7.3.2.20.2.1.11, 7.6.2.5.5, TS3.3, ORM Att. 2 Regulatory Guide 1.106, Rev. 1 (March 1977)

Thermal Overload Protection for Electric Motors on Motor-Operated Valves Project Position - Comply with the following clarification:

Reference:

Paragraph c.1(a) - CPS complies with the intent of this regulatory guide in that thermal overload protection is continuously bypassed with the exception that it may be placed into service for short periods of time during valve maintenance, testing, and repositioning during normal plant operation. This

action is limited by the requirements of ORM 2.5.2. Therefore, automatic actuation devices to bypass the thermal overloads during accident conditions are not required. As such, the CPS program for control of MOV thermal overload protection devices complies with the requirements of Regulatory Guide 1.106 to

prevent overload protection circuitry from inhibiting the ability of MOVs to perform their safety function. USAR Section - 8.1.6.1.19; ORM 2.5.2, 3.5.2, 4.5.2.1, 4.5.2.2, and 5.5.2.

Regulatory Guide 1.107, Rev. 1 (February 1977)

Qualifications for Cement Grouting for Prestressing Tendons in Containment Structures Project Position - Not applicable since Clinton Power Station does not have prestressed tendons in its containment structures.

CPS/USAR CHAPTER 01 1.8-53 REV. 11, JANUARY 2005 Regulatory Guide 1.108, Rev. 1 (August 1977)

Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants Project Position - Comply with the following clarifications: 1)

Reference:

Section C.1.b(3) - the Division 3 diesel generator design includes override capability to ensure automatic switchover from the test mode to ready-to-load operation in response to a loss-of-coolant-accident (LOCA) initiation signal. However, during testing with a non-zero droop setting in effect (to support paralleling the diesel generator with the offsite power source), in the event of a LOCA initiation signal concurrent with a loss of the offsite power source to the bus, operator action may be required (in addition to the automatic actions) to reset the governor and thus ensure bus frequency is within required limits when the diesel generator alone is subsequently supplying power to the Division 3 bus. 2)

Reference:

Section C.2 - Periodic testing of diesel generator units will be as required by the Technical Specifications. Reporting of diesel generator unit test failures will be as required by the Operational Requirements Manual. 3)

Reference:

Section C.2.c - Tests, such as the largest load rejection test, full load rejection test, 24-hour run, and the test mode override test, may be performed during normal plant operations, as well as during plant shutdown. (Reference

License Amendment 132). USAR Section/Subsection - 8.3.1; TS 3.8.1, 3.8.2; ORM 6.9.2.1 Regulatory Guide 1.109, Rev. 1 (October 1977)

Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I Project Position - Comply with the following clarification:

Reference:

Paragraph C.2 - The effect of a finite cloud and elevated plume will be considered for all releases meeting the criteria of Regulatory Guide 1.111, Revision 1, Paragraph C.2.b. This effect will not be limited to stacks more than 80 meters high. USAR Sections - 11.2, 11.3, 15.6.5.5.1, TS 5.6.3, and ODCM CPS/USAR CHAPTER 01 1.8-54 REV. 11, JANUARY 2005 Regulatory Guide 1.110, Rev. 0 (March 1976)

Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors Project Position - Clinton Power Station has opted to comply with Annex to 10 CFR 50, Appendix I. Hence, this regulatory guide is not applicable.

Regulatory Guide 1.111, Rev. 1 (July 1977)

Methods for Estimating Atmospheric Transports and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors Project Position - Comply with the following clarification:

Reference:

Paragraph C.3.a - Radio-decay will be considered individually for each nuclide. "Conservative" estimates considered here are unnecessary. USAR Sections - 2.3, 11.3, ODCM Regulatory Guide 1.112, Rev. O-R (May 1977)

Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors Project Position - Comply USAR Subsections - 11.2.3, 11.3.3 Regulatory Guide 1.113, Rev. 1 (April 1977)

Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I Project Position - Comply USAR Subsections - 2.4.12, 11.2.3 CPS/USAR CHAPTER 01 1.8-55 REV. 14, JANUARY 2011 Regulatory Guide 1.114, Rev. 2 (May 1989)

Guidance on Being Operator at the Controls of a Nuclear Power Plant Project Position - Comply with the following clarification:

Reference:

Footnote 2 - This footnote defines operational control panels as those that enable the operator at the controls to perform required manual safety functions and equipment surveillance and to monitor plant conditions under normal and accident conditions. Unobstructed view of and access to these panels is required. The nature of the CPS design is such that certain surveillance and monitoring actions not requiring prompt corrective action will be conducted in back row panels. USAR Section - 13.5 Regulatory Guide 1.115, Rev. 1 (July, 1977)

Protection Against Low-Trajectory Turbine Missiles Project Position - Comply. USAR Subsection - 3.5.1.3 Regulatory Guide 1.116, Rev. O-R (June 1976)

Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems Project Position - Comply with the following clarification: As committed to in the PSAR, Installation, Inspection and Testing of Mechanical Equipment and Systems during the design and construction of CPS was in compliance with ANSI N45.2.8 (Draft 3, Rev. 3 April, 1974), as endorsed by WASH 1309, UC-80.

The draft standard was replaced by ANSI N45.2.8 - 1975 of the subject Regulatory

Guide. Superceded by ASME, NQA-1 (1994) Subpart 2.8. USAR Chapters - 14, 17 CPS/USAR CHAPTER 01 1.8-56 REV. 11, JANUARY 2005 Regulatory Guide 1.117, Rev. 1 (April, 1978)

Tornado Design Classification Project Position - The project complies with the requirements of Regulatory Guide 1.117 with the following clarifications: The discussion contained in Regulatory Guide 1.117 states that protection of designated structures, systems, and components may generally be accomplished by designing protective barriers to preclude tornado damage, and if protective barriers are not installed, the structures and components themselves should designed to withstand the effects of the tornado, including tornado missile strikes.

Important systems and components (as discussed in Regulatory Guide 1.117) are generally protected. The limited amount of unprotected portions of important systems and components are analyzed using a probabilistic missile strike analysis consistent with the acceptance criteria in Standard Review Plan 3.5.1.4, Missiles Generated By Natural

Phenomena. USAR Subsection - 3.5.1.4, 3.5.2.4, and 3.5.2.5 Regulatory Guide 1.118, Rev. 2 (June 1978)

Periodic Testing of Electric Power and Protection Systems Project Position - The Project complies with NRC Regulatory Guide 1.118 with the following clarifications: 1)

Reference:

Paragraph C.6 - Trip of an associated protective channel or actuation of an associated Class 1E Load Group is required on removal of fuses or opening of a breaker only for the purpose of deactivating instrumentation or control circuitry. 2)

Reference:

Paragraph C.8.a - Appropriate state-of-the art technology will be implemented to periodically assure that the sensor (beginning at the sensor input), trip unit, logic, and actuator response times have not deteriorated so as to compromise the respective system design requirements. 3)

Reference:

Paragraph C.8.b - Test intervals, both initial and revised, should be such that significant changes in failure rates can be detected. The Nuclear System Protection System has been designed to support the requirements of this guide. USAR Chapters - 7 and 8 CPS/USAR CHAPTER 01 1.8-57 REV. 11, JANUARY 2005 Regulatory Guide 1.119 has been withdrawn Regulatory Guide 1.120, Rev. 1 (November 1977)

Fire Protection Guidelines for Nuclear Power Plants Project Position - The fire protection guidelines for the Clinton Power Station were taken from the Branch Technical Position APCSB 9.5-1 Appendix A, "Guidelines for Fire Protection for Nuclear Power Plants Docketed prior to July 1, 1976". A complete evaluation of the projects' compliance with this Branch Technical Position is contained in Section 4.0 of the Clinton Power Station Fire Protection Evaluation Report.

Reference - Appendix E Regulatory Guide 1.121, Rev. 0 (August 1976)

Bases for Plugging Degraded PWR Steam Generator Tubes Project Position - Not applicable to BWRs.

Regulatory Guide 1.122, Rev. 1 (February 1978)

Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components Project Position - This Regulatory Guide was first issued in September, 1976 while the date of construction permit for Clinton is February, 1976. However, the project complies with the intent of the Regulatory Guide 1.122 with the following clarification: Floor response spectra were generated at fifty periods lying between .02 to 2.0 seconds interval. These periods compare well with the recommended values of periods in Table 1 of the Regulatory Guide 1.122 and therefore the selected periods in Clinton project for floor response spectra generation are satisfying the intent of the Regulatory Guide 1.122. USAR Subsections - 3.7.2.5, 3.7.2.5.1, 3.7.2.5.2, 3.7.2.5.3 CPS/USAR CHAPTER 01 1.8-58 REV. 11, JANUARY 2005 Regulatory Guide 1.124, Rev. 1 (January, 1978)

Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports Project Position - Comply. USAR Subsection - 3.9.3 Regulatory Guide 1.125, Rev. 1 (October, 1978)

Physical Models for Design and Operation of Hydraulic Structures and Systems for Nuclear Power Plants Project Position - Physical hydraulic models were not used for hydraulic design of Clinton Power Station structures, therefore, Regulatory Guide 1.125 is not applicable.

Regulatory Guide 1.126, Rev. 1 (March 1978)

An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification Project Position - GE meets the requirements of this regulatory guide with the exception of the method of derivation of the densification values for resintering tests.

GE believes that the NRC method in the regulatory guide represents a significant departure from previously approved GE methods. Specifically, this departure requires an increase in the level of statistical confidence level from 50 percent to 95 percent employed to infer population parameters from the sampling results. This departure from the GE approach previously approved by the Staff is interpreted as the introduction of additional conservatism in the GE densification analyses, and such a Staff position is considered inappropriate in recognition of the GE/NRC agreed upon conservatisms incorporated into the individual densification models during their

formulation. USAR Section - 4.2 CPS/USAR CHAPTER 01 1.8-59 REV. 11, JANUARY 2005 Regulatory Guide 1.127, Rev. 1 (March 1978)

Inspection of Water-Control Structures Associated with Nuclear Power Plants Project Position - The requirements of Regulatory Guide 1.127 are not specifically applicable to the Lake Clinton Dam since the dam is not required for the emergency cooling water system or flood protection of the Clinton Power Station. The requirements of Regulatory Guide 1.127 are, however , applicable to the submerged dam and baffle dike within the lake that forms the Ultimate Heat Sink. CPS complies with the Regulatory Guide 1.127 in respect to the submerged dam and baffle dike. USAR Subsections - 2.4.11.6 and 2.5.6.8 Regulatory Guide 1.128, Rev. 1 (October, 1978)

Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants Project Position - The project complies with the NRC Regulatory Guide 1.128 with the following exceptions and clarification:

Exceptions: 1)

Reference:

Paragraph C.1 - CPS complies with the IEEE 484 requirement of limiting the hydrogen accumulation to less than two percent of the total battery area volume. The regulatory position of limiting the concentration to less than two percent at any location within the battery area would be impossible to verify. 2)

Reference:

Paragraphs C.4 & C.6.h. IEEE Standard 450-1995 revision is used in lieu of earlier revisions. Clarification:

Reference:

Paragraph C.2 - Compliance to Regulatory Guide 1.120 is addressed in USAR Subsection 9.5.1 and the Fire Protection Evaluation Report for Clinton Power Station. USAR Section/Subsection - 8.1, 9.5.1, Appendix E.

CPS/USAR CHAPTER 01 1.8-60 REV. 14, JANUARY 2011 Regulatory Guide 1.129, Rev. 1 (February 1978)

Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants Project Position - Conformance with this Regulatory Guide is provided by the maintenance and testing program described in the CPS Technical Specifications: 1)

Reference:

Paragraphs B,C, & C.2. IEEE Standard 450-1995, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications, is used in lieu of earlier revisions. 2)

Reference:

Paragraph C.1. Battery performance discharge tests are performed as required in CPS Technical Specifications 3.8. USAR Section - TS 3.8.4, TS 3.8.5, TS 3.8.6, USAR 8.3.2 Regulatory Guide 1.130, Rev. 1 (October 1978)

Service Limits and Loading Combinations for Class 1 Plate - and - Shell-Type Component Supports Project Position - Comply USAR Subsections - 3.8.3.5.5, 3.9.3 CPS/USAR CHAPTER 01 1.8-61 REV. 11, JANUARY 2005 Regulatory Guide 1.131, Rev. 0 (August 1977)

Qualification Tests of Electric Cables, Field Splices and Connections for Light-Water-Cooled Nuclear Power Plants Project Position - Comply with respect to testing of cables. See section 8.1.6.24 for statements concerning qualification testing of UCI splicing tape. USAR Subsections - 8.1.6 CPS/USAR CHAPTER 01 1.8-62 REV. 11, JANUARY 2005 Regulatory Guide 1.132 Rev. 1 (March, 1979)

Site Investigations for Foundation of Nuclear Power Plants Project Position - Most of the geotechnical site investigation work for Clinton Power Station was done prior to the original issuance of Reg. Guide 1.132 in September, 1977, therefore compliance was not possible. However, the work done complies with Reg.

Guide 1.132 with the following exceptions: (1)

Reference:

Paragraph C.2 - The coordinates for the borings are not shown on the boring logs. However, drawings showing the locations of these borings are

provided and have the state plane coordi nate system identified on them. (2)

Reference:

Paragraph C.3 - Piezometers were not used to monitor the effects of dewatering for the main plant excavation. The principle reasons for not establishing a monitoring system were: 1. The surrounding tills were impermeable and very little seepage was expected and obtained. 2. The major dewatering problem would be the removal of rain water.

3. There were no ground water users in the area of the excavation that would be affected. (3)

Reference:

Paragraph C.5 - The boring programs for the Main Plant complex, the Emergency Core Cooling System (ECCS) pipeline, the Ultimate Heat Sink, and the Main Dam do not fully comply with the requirements as provided in Appendix C of this regulatory guide. However, the borings for the Main Plant Ultimate Heat Sink, and Main Dam were located and drilled to provide an adequate geologic cross-section of the respective areas. It was also felt that the borings located near the Main Plant provided an adequate description of the subsurface materials to be encountered along the ECCS pipeline. Therefore, additional borings along the pipeline route were not drilled. (4)

Reference:

Paragraph C.6 - Continuous sampling of the soils encountered in the borings was performed in a few borings, but not to the extent required by this regulatory guide. As previously stated, the boring and sampling programs were established to provide adequate geologic information to design the structures. USAR Subsections - 2.5.4, 2.5.5, 2.5.6 CPS/USAR CHAPTER 01 1.8-63 REV. 11, JANUARY 2005 Regulatory Guide 1.133, Rev. 1 (May 1981)

Loose Part Detection Program for the Primary System of Light-Water-Cooled Reactors Project Position - This Regulaotry Guide is no longer a requirement for BWRs as accepted by NRC SER contained within NEDC-32975P-A February 2001. USAR Subsection - N/A Regulatory Guide 1.134, Rev. 2 (April 1987)

Medical Certification and Monitoring of Personnel Requiring Operating License

Project Position - Comply USAR Section/Subsection - 13.1 Regulatory Guide 1.135 Rev. 0 (September, 1977)

Normal Water Level and Discharge at Nuclear Power Plants Project Position - Comply. USAR Section - 2.4 Regulatory Guide 1.136, Rev. 1 (October, 1978)

Material for Concrete Containments (Article CC-2000 of the "Code for Concrete Reactor Vessels and Containments")

Project Position - This Regulatory Guide and the referenced Code, 1977 Edition, did not exist at the time the Construction Permit was obtained. Nevertheless, the project is not in conflict with the regulatory positions C.1 and C.2. Regulatory positions C.3 and C.4 do not apply. USAR Reference - Appendix B CPS/USAR CHAPTER 01 1.8-64 REV. 15, JANUARY 2013 Regulatory Guide 1.137, Rev. 0 (January 1978)

Fuel Oil Systems for Standby Diesel Generators Project Position - The Project complies with NRC Regulatory Guide 1.137 with the following clarifications and exceptions: (1)

Reference:

ANSI N195-1976 Section 4, Paragraph (3) - The Diesel Fuel Oil System is located in the plant vital area except for the last portion of the fill and vent lines which extend outside the building. (2)

Reference:

ANSI N195-1976 Section 5.4 - Stored fuel requirements for emergency diesel generators (EDGs) at CPS are determined using post-LOCA maximum electrical load demands for each EDG for 7 days versus the continuous rating or time-dependent diesel loading of each EDG with 10%

margin. (3)

Reference:

ANSI N195-1976 Section 6.3 and Appendix A - A duplex strainer is not provided in the transfer pump suction line to preclude a postulated loss of pump suction by strainer plugging. A Y-type strainer is provided in a recirculation line. The tank outlet nozzle is six inches above the bottom of the tank to prevent sludge from entering the line. (4)

Reference:

ANSI N195-1976 Section 7.5 Paragraph 1 - A non-nuclear safety related strainer is provided in each fill line. There is a locked closed valve between the strainer and the storage tank to preclude entry of deleterious material. (5)

Reference:

ANSI N195-1976 Section 8, Paragraph 1d - A high level alarm is not provided for the supply tanks. A high level alarm for the storage tank could be useful only when filling the tank. Since there are level indications which will be monitored during the filling, a high level alarm would not enhance the safety of the system. (6) The fuel oil will be sampled and analyzed periodically to verify that its quality meets the diesel manufacturer's recommendations, the requirements of ASTM-D975-06b and the CPS Technical Specifications. The methods and tests specified in ASTM-D975 will be followed when there are differences between ASTM-D975 and the diesel manufacturer's recommendations. However, all the diesel manufacturer's recommended limits will be met when they are relevant to the test methods specified in ASTM-D975. (7)

Reference:

Paragraph C.1.g The Diesel Fuel Oil fill lines have a protective coating and an impressed type cathodic protection system, although the cathodic protection system may not meet the requirements of NACE Standard RP-01-69.

This is adequate because the fill lines are not required for proper operation of the diesel generator units during post-LOCA maximum load demands. All other equipment in the Diesel Generator Fuel-Oil Storage and Transfer System do meet NACE Standard RP-01-69 requirements.

CPS/USAR CHAPTER 01 1.8-64a REV. 15, JANUARY 2013 (8) Fuel oil will be sampled in accordance with ASTM D4057-95 (9)

Reference:

Paragraph C.2.f. The current 25% interval extension that is allowed by Technical Specification Surveillance Requirement (SR) 3.0.2 for the 10-year tank cleaning SR 3.8.3.6 removed by License Amendment 186 is transferred to this Section. The Bases of SR 3.0.2 provide the caution that this 25% extension is not intended to be used repeatedly merely as an operational convenience. USAR Subsection - 9.5.4; TS 3.8.3 Q&R 040.33, Q&R 040.34 CPS/USAR CHAPTER 01 1.8-65 REV. 11, JANUARY 2005 Regulatory Guide 1.138, Rev. 0 (April, 1978)

Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Plants Project Position - Comply. USAR Subsections - 2.5.4, 2.5.5, 2.5.6 Regulatory Guide 1.139, For Comment (May 1978)

Guidance for Residual Heat Removal The Clinton plant is designed so that the RHR system complies with the intent of this guide except as follows: 1. Regulatory Position C.2.a, second sentence suggests alarms which the RHR system does not have. 2. Regulatory Position C.2.a, third and fifth sentences suggest independent and diverse interlocks. The RHR interlocks are not diverse. 3. Regulatory Position C.2.a, fourth sentence suggests that failure of a power supply should not cause any valve to change position. Failure of the Reactor Protection System instrument power bus in one division will cause automatic isolation of the RHR shutdown suct ion valve in that division. The Clinton design complies with Regulatory Guide 1.139, by using the following alternate approaches to Regulatory Guide 1.139 position C-2a and C-5. 1. Position C.2.a, second sentence - The design provides for valve position indication in the control room. The addition of alarms to alert the operator that an RHR valve is open is not considered necessary because of the number of procedural and administrative controls that must be exercised in taking the reactor from a power mode to the RHR mode and because the reactor high pressure signals are permissively interlocked into the valve opening logic to disallow valve opening on reactor high pressure. 2. Position C.2.a, third and fifth sentences - The design complies with these suggestions by the use of independent interlocks and series valves powered by redundant power sources. 3. Position C.2.a, fourth sentence - The design complies with the intent of the Regulatory Guide (i.e., protect the low pressure piping). Upon failure of logic the RPS power supply, valves are signaled to close. However, division power would need to be available to implement the change in valve position to the close position. USAR Subsection - 5.4.7 CPS/USAR CHAPTER 01 1.8-66 REV. 11, JANUARY 2005 Regulatory Guide 1.140, Rev. 0 (March 1978)

Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants Project Position - All of the equipment of the filter systems were specified and purchased prior to the issuance of Regulatory Guide 1.140. However, the design of the non-safety-related filter systems meet the intent of the requirements of this guide, except as noted

below: (1)

Reference:

C.1.a and C.1.b - The equipment and components (excluding charcoal and filter pads) are designed to withstand a maximum of 40 year integrated radiation dose and worst-case anticipated continuous service, rather than 40 years of continuous service. (2)

Reference:

C.2.a - All of the exhaust systems contain ducts, dampers, fans, related instrumentation, prefilters and HEPA filters. Charcoal adsorbers are only used when iodine is anticipated to be present. Heaters are only used upstream of charcoal adsorbers when the potential exist for the air stream relative humidity to exceed 70%. (3)

Reference:

C.2.d - For discussion of conformance to Regulatory Guide 8.8, see discussion under Regulatory Guide 8.8 in this section. (4)

Reference:

C.2.3 - The exhaust systems are not directly connected to outdoors because they take their suction from inplant areas. Therefore, this section is not applicable. (5)

Reference:

C.2.f - All filter housings are of welded construction. All the filter housings and associated ductwork are designed to meet the requirements of Section 4.12 of ANSI N509-1976, but were not leak tested. (6)

Reference:

C.3.b - The HEPA filters are designed and constructed in accordance with ANSI N509-1976, Section 5.1. The filter banks are field tested

in accordance with ANSI N510-1980. Further procurement of HEPA filters after January 1, 1986 shall be in accordance with ANSI N509-1980. (7)

Reference:

C.3.e - Even though prefilter and HEPA filter bank in 0VW06SA/SB is higher than three HEPA filters and no permanent gallery is provided, the access areas are adequate to support servicing and maintenance of filters and

lighting. (8)

Reference:

C.3.f - Ductwork associated with non-safety related filter systems is not designed to post-LOCA pipe break loadings, or to exhaust wind conditions such as tornadoes. However, all ductwork in Seismic Category I buildings is

seismically supported. (9)

Reference:

C.3.h - The drywell purge charcoal adsorber shall be field leak tested in accordance with ANSI N510-1980. Viewports are provided to observe adequate fill of the charcoal adsorber.

CPS/USAR Regulatory Guide 1.140, Rev. 0 (March 1978) (Cont'd) CHAPTER 01 1.8-67 REV. 11, JANUARY 2005 (10)

Reference:

C.3.i - The system fans and motors, mounting, and ductwork connections are designed, constructed and tested in accordance with the intent of ANSI N509-1976. (11)

Reference:

C.3.l - The dampers were designed in accordance with the intent of ANSI N509-1976. (12)

Reference:

C.4.b - A clearance of approximately two feet, instead of three feet, has been provided between the upstream edges of the HEPA filter mounting frames and the adjacent upstream components of filter units 0VQ01SA/SB/SC. This is acceptable since a minimum 4 feet 7 inch clearance has been provided for maintenance of these filters from the downstream side. (13)

Reference:

C.6.a - The activated carbon is in accordance with Regulatory Guide 1.140, except that it is tested to the requirements of ANSI N509-1980, Table 5-1. (14) In addition, the CPS "ANSI N509/510 Variance Report," HVAC-02-CP identifies minor deviations to this Regulatory Guide referenced ANSI standards. This Variance Report is a controlled document, maintained by the Nuclear Station Engineering Department. USAR Subsections - 9.4.7.2, 9.4.9, 9.4.11, 9.4.13 Regulatory Guide 1.141, Rev. 0 (April 1978)

Containment Isolation Provisions for Fluid Systems Project Position - Comply with the following clarification: 1. Reference ANSI N271-1976 Paragraph 4.2.3 - A short circuit in the valve indication circuit would cause the fuse at the motor control center to open, thus rendering the valve electrically inoperable. This is an acceptable condition since only a single failure in one Electrical Division need be considered. Inboard and outboard isolation valves have different Divisional assignments so that containment isolation can be accomplished even considering the loss of one of the two valves. 2. Reference Regulatory Guide Position C.2 - Valve Numbers lE12-F008, 1CC071, 1CC072, ICC073, 1CC074, 1CY020, 1CY021, 1FP051, 1FP054, 1FP078 and 1FP079 will have power removed by locking the circuit breaker in the open position. During the time that power is removed from the valve it will be considered as a manual valve and will not have position indication in the Main Control Room. During times that power is applied and the valve is required to be open, valve position indication will be provided. USAR Section - 6.2.6, TS3.6.1.3 CPS/USAR CHAPTER 01 1.8-68 REV. 11, JANUARY 2005 Regulatory Guide 1.142, Rev. 0 (April, 1978)

Safety Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments)

Project Position - Comply with the following clarifications: 1. R.G. 1.142 generally endorses ACI 349-76. However, the Regulatory Guide does not endorse the provisions of ACI 349-76 as adequate for drywell structures. CPS design utilizes the provisions of Sections 10.6.3 of ACI 349-80 in place of the 1976 version of the standard. 2. In some cases, CPS utilized the bend test requirements of ASTM A615 in place of ACI 349. Additional detail on this clarification is provided in USAR Section 3.8.3.2.2. USAR Subsections - 3.8.3, 3.8.4, 3.8.5.

Regulatory Guide 1.143, Rev. 0 (July 1978)

Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water Cooled Nuclear Power Plants Project Position - The Project complies with this guide with the following exceptions and clarifications: 1.

Reference:

Section B, Page 2, First Paragraph, First Sentence - It is clarified that the radwaste systems for CPS are considered to begin with and include the interface valves. 2.

Reference:

Section B, Page 2, First Paragraph, Second Sentence - It is clarified that the radwaste system for CPS terminates at the end of the pipe containing the last isolation valve before the point of controlled discharge to the service water discharge line, or at the end of the pipe containing the last isolation valve before the cycle condensate storage tank, or at the point of storage of packaged solid waste prior to shipment offsite to licensed burial ground. 3.

Reference:

Paragraphs C.1.1.2, C.2.1.2, C.3.1.2 Materials for pressure-retaining components conform to the requirements of ASTM Specifications. 4.

Reference:

Paragraph 4.3 - It is clarified that the scope of radwaste system pressure testing includes all pressure-retaining components, appurtenances, and completed systems. Bolts, studs, washers, gaskets, and possible localized instances of pump and valve packing are exempted from the pressure test. This is consistent with ASME Section III NB6000 and ANSI B31.1 (1983 edition). The

Off-Gas System will be pneumatically tested at a minimum of 75 psig for no less than 30 minutes.

CPS/USAR CHAPTER 01 1.8-69 REV. 11, JANUARY 2005 Portions of the radioactive waste management system contain polypropylene lined steel pipe and valves. These pipes and valves are in the demineralizer subsystem because of the superior corrosion resistance of polypropylene to chemicals. This portion of the subsystem will only be inservice leak tested at normal operating pressure. 5. Non-Category 1 equipment is evaluated per DC-ME-17-CP, Revision 1.

6. Plastic spacers are used on an as needed basis at flanged joints of plastic lined pipe. The spacers are provided by the manufacturer of the lined piping to adjust for small fit-up differences. 7.

Reference:

Paragraph C.1.2.1 - There are no local alarms to alert people in the area of potential overflow conditions. In lieu of this, the operator annunciator response procedure in the Radwaste Operations Center (ROC) has a required

action to make a plant wide announcement regarding the potential overflow and alerting individuals in the area of potentially changing radiological conditions. 8.

Reference:

C.3.1 - Materials for non-pressure retaining attachments and appurtenances to solid waste tanks are not required to be constructed of materials conforming to the requirements of Section II of the ASME Boiler and Pressure Vessel Code. Examples of this type of equipment are nozzles, piping and welds in atmospheric tanks which do not carry process fluid and whose elevation is above the tank overflow level. In addition, these non-pressure retaining attachments and appurtenances are exempted from the design, construction and testing criteria set forth in this section, section C.4 and Table 1 to this regulatory guide. Failure of these components would not result in a radioactive release. USAR Sections - 11.2.1, 11.3.2.2.1, 11.4.2, 11.5.1.1.2

CPS/USAR CHAPTER 01 1.8-70 REV. 15, JANUARY 2013 Regulatory Guide 1.147 Inservice Inspection Code Case Acceptability ASME Section XI Division 1 Project Position - Comply with the following understandings: 1. Application to the Commission for acceptance of selected code cases issued after the latest revision of the guide may be made. 2. Acceptable code cases annulled by action of the ASME council (or deleted in later revisions to this guide), but specified in the Clinton Power Station Preservice, or Inservice, Inspection Program, as applicable, shall remain valid. Commitment to meet a specific revision of this guide has little significance since the guide is revised as new code cases are issued or deleted by the ASME and its actions approved by the NRC. USAR Subsection - 5.2.4, 6.6.

Regulatory Guide 1.149 Revision 4 (April 2011)

Nuclear Power Plant Simulation Facilities For Use In Operator Training, License Examinations, and Applicant Experience Requirements Project Position - Comply USAR Subsection - 13.2.1 CPS/USAR CHAPTER 01 1.8-71 REV. 15, JANUARY 2013 This Page Intentionally Left Blank.

CPS/USAR CHAPTER 01 1.8-72 REV. 15, JANUARY 2013 This Page Intentionally Left Blank.

CPS/USAR CHAPTER 01 1.8-73 REV. 15, JANUARY 2013 Regulatory Guide 1.150, Rev. 1 (February 1983)

Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations Project Position - Comply with alternative methods as described in Appendix A of this Guide. USAR Subsection - 5.2.4 CPS/USAR CHAPTER 01 1.8-74 REV. 14, JANUARY 2011 Regulatory Guide, 1.155 (August 1988)

Station Blackout Project Position - Illinois Power submitted its compliance to the Station Blackout Rule, 10CFR50.63, "Loss of All Alternating Current Power," in letters to the NRC dated April 16, 1989; March 30, 1990; May 17, 1990; July 6, 1992; October 29, 1992; and December 22, 1992. USAR Subsection - Table 8.1-3 Regulatory Guide 1.163, Rev. 0 (September 1995)

Performance-Based Containment Leak-Test Program Project Position - Comply with the following clarification: Bechtel Topical Report BN-TOP-1 is al so an acceptable option for performance of Type A tests. USAR Subsection - 6.2.6 Regulatory Guide 1.181, Rev. 0, (September 1999)

Content of the Update Final Safety Analysis Report in Accordance with 10 CFR 50.71(e)

Project Position - Comply USAR Subsection - 1.1.9.2, 1.1.9.7 Regulatory Guide 1.182, Rev. 0 (May 2000)

Assessing and Managing Risks Before Maintenance Activities at Nuclear Power Plants Project Position - Comply USAR Subsection - 16 Regulatory Guide 1.183, Rev. 0 (July 2000)

Alternative Radiological Source Terms for Evaluating Design Basis Accidents At Nuclear Power Reactors Project Position

- Comply USAR Subsection - 15.4.9, 15.6.4, 15.6.5, 15.7.4 CPS/USAR CHAPTER 01 1.8-74a REV. 14, JANUARY 2011 Regulatory Guide 1.196, Rev. 0 (May 2003)

Control Room Habitability At Light-Water Nuclear Power Reactors Project Position - The project only complies with Section C. 2.7.3, Degraded and Nonconforming Conditions of Regulatory Guide 1.196 Revision 0. The Control Room Envelope Habitability Program is governed by Technical Specification 5.5.15 approved under License Amendment No. 178. Technical Specification Program 5.5.15, "Control Room Envelope Habitability Program" is a result of a commitment of the response to NRC Generic Letter 2003-01 by implementing the requirements of the Consolidated Line Item Improvement Process for implementation of TSTF-448 Revision 3. The Control Room Envelope Habitability Program was created as a result of findings at facilities that existing Technical Specifications may not be adequate to ensure the requirements of 10CFR50 Appendix A GDC 19 are met as described in Generic Letter 2003-01. Survey of chemical sources is to be performed at least once per 6 years as part of the Periodic Assessment of the Control Room Envelope Habitability Program. UFSAR Section - 6.4 and 9.4.1 Regulatory Guide 1.197, Rev. 0 (May 2003)

Demonstrating Control Room Envelope Integrity At Nuclear Power Reactors Project Position - The project has only committed to the requirements for determining the unfiltered air inleakage past the Control Room Envelope boundary into the Control Room Envelope is in accordance with the testing methods and at the frequencies specified in Section C.1 and C.2 of Regulatory Guide 1.197, Revision 0 as described by Technical Specification 5.5.15, Control Room Envelope Habitability Program. USAR Section - 6.4 and 9.4.1 Regulatory Guide 4.1, Rev. 1 (April 1975)

Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants Project Position - Comply USAR Subsection - 12.5.2 CPS/USAR CHAPTER 01 1.8-75 REV. 14, JANUARY 2011 Regulatory Guide 4.13, Rev. 1 (July 1977)

Performance, Testing and Procedural Specifications for Thermoluminescent Dosimetry: Environmental applications Project Position - Comply ER Subsection

6.2.5, USAR Subsection: 12.5.2 Regulatory Guide 4.15, Rev. 1 (February 1979)

Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment Project Position - Comply USAR Subsection - 12.5.2, ORM 6.5.2.8 Regulatory Guide 5.66, Rev 0 (June 1991)

Access Authorization Program for Nuclear Power Plants Project Postion - Comply USAR Subsection - 13.1.1.1.3, 13.6 Regulatory Guide 8.1, Rev. 0 (February 1973)

Radiation Symbol Project Position - Comply USAR Subsection - 12.5.2 CPS/USAR CHAPTER 01 1.8-76 REV. 14, JANUARY 2011 Regulatory Guide 8.2, Rev. 0 (February 1973)

Guide for Administrative Practices in Radiation Monitoring Project Position - Comply with the following exception: Licensee and contractor personnel will be processed to become radiation workers at Clinton Power Station without a radiation worker physical.

All Exelon personnel will receive a pre-employment physical equivalent to a Radiation Worker physical. The industrial risks are not any higher than those experienced at fossil plants where contractor physicals are not provided. Additionally, a proficiently managed ALARA program should minimize the risk of potential litigation resulting from routine exposure to ionizing radiation.

Reference:

ANSI N13.2-1969, Paragraph 4.7.1 USAR Subsections - 12.3.4.4.1, 12.5.2, 13.2.3 Regulatory Guide 8.3, Rev. 0 (February 1973)

Film Badge Performance Criteria Project Position - This Regulatory Guide is not applicable at Clinton Power Station. Clinton Power Station uses Thermoluminescent Dosimeters in accordance with

10CFR20.1501(c).

Regulatory Guide 8.4, Rev. 0 (February 1973)

Direct Reading and Indirect - Reading Pocket Dosimeters Project Position - Comply with the following exception: 1.

Reference:

Paragraph C.3 - In mixed radiation fields, more suitable and accurate dosimeters have been placed in service to determine neutron and gamma exposures; therefore, pocket dosimeters will not be used for this purpose. USAR Subsection - 12.5.2 CPS/USAR CHAPTER 01 1.8-77 REV. 14, JANUARY 2011 Regulatory Guide 8.5, Rev. 1 (March 1981)

Criticality and Other Interior Evacuation Signals Project Position - The project complies with the following exception: 1) The NRC granted an exemption to the licensee of Clinton Power Station concerning 10CFR70.24 and the requirement to have a criticality alarm system in the fuel storage area (see pages 5 and 6 of Operating License NPF-62). USAR Subsections - 9.5.2, 12.3.4, 12.5.2 Regulatory Guide 8.6, Rev. 0 (May 1973)

Standard Test Procedure for Geiger - Mueller Counters Project Position - The project complies with Regulatory Guide 8.6 with the following clarification: Operational testing and calibration of Geiger - Mueller type detectors is performed in accordance with American National Standards N323-1978 "Radiation Protection Instrumentation Test and Calibration," Sections 4.1, 4.2.2.1, 4.5, and 5.1. Source check frequency requirements shall be established by procedure. Source check acceptance ranges shall be consistent with applicable portions of ANSI N323-1978 or ANSI N320-1979, "Performance Specifications for Reactor Emergency R adiological Monitoring Instrumentation," depending on the intended application of the instrument. USAR Subsections - 12.5.2, 12.5.3 Regulatory Guide 8.7, Rev. 1 (June 1992)

Instructions for Recording and Reporting Occupational Radiation Exposure Data Project Position - Comply USAR Subsection - 12.5.2 CPS/USAR CHAPTER 01 1.8-78 REV. 14, JANUARY 2011 Regulatory Guide 8.8, Rev. 3 (June 1978)

Information Relevant to insuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as Reasonably Achievable Project Position - The project complies with Regulatory Guide 8.8 with the following clarifications and exceptions: 1)

Reference:

C.1.b(3) - NRC staff is considering certification by peer qualifications for RPM candidates. Should this be adopted, Clinton Power Station will take exceptions to this portion of regulatory guide. 2)

Reference:

C.2.a - Clinton Power Station's request concerning an exception to 10CFR 20.1601 to raise the radiation levels requiring locking to 1000 mrem/hr as opposed to the presently stated level of 100 mrem/hr was approved. Administrative controls will provide the effective control over ingress to areas greater than 100 mrem/hr. 3)

Reference:

C.2.c.(4) - The Clinton Project complies with this position in so far as practical. Wherever pressure gauges rather than transmitters are used, they are back flushed with clean water so as to reduce the potential for exposure at the gauge readout locations. 4)

Reference:

C.2.d.(3) - Temporary openings to exhaust ducts for local control of airborne contaminants during equipment maintenance are not provided, since use of these may result in imbalance of the exhaust system. Imbalance of the exhaust system may result in loss of airborne contamination control in adjacent areas. 5)

Reference:

C.2.g (2) (Ref. 10) - ANS/HPS 56.8 "Location and Design Criteria for Area Radiation Monitoring Systems for LWR" (draft) is identified by the Regulatory Guide as the reference for placement of detectors for optimum coverage of areas, however that document was never issued. Clinton Power Station will use ANSI/ANS-HPSSC 6.8.1-1981 "Location and Design Criteria for Area Radiation Monitoring Systems for Light Water Nuclear Reactors" with the following exception. Table 1 lists examples of locations for area radiation monitors; Clinton Power Station's placement and function is described in Table

12.3-2. 6)

Reference:

C.2.h.(5) - Where use of tees cannot be avoided, tees are oriented in the branch horizontally or above the run as allowed by physical constraints. 7)

Reference:

C.2.(h) - ANS N197 and AN S 55.1 were not specifically considered in the Clinton Power Station Design. In addition, CPS has committed to and will comply with the requirements of Regulatory Guide 8.8 (Revision 4) C.4.d(1) and (2). (Q&R 471:17) USAR Section/Subsection - 12.1.2, 12.3 , 12.5.2, 13.2.3 CPS/USAR CHAPTER 01 1.8-79 REV. 14, JANUARY 2011 Regulatory Guide 8.9, Rev. 1 (July 1993)

Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program Project Position - Comply USAR Subsections - 12.3.1, 12.5.2 Regulatory Guide 8.10, Rev. l-R (September 1975)

Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable Project Position - Comply USAR Subsections - 12.1.3, 12.5.2, 13.2.3 Regulatory Guide 8.11, Rev. 0 (June 1974)

Applications of Bioassay for Uranium Project Position - This guide does not apply to Clinton Power Station.

Regulatory Guide 8.12, Rev. 0 (December 1974)

Criticality Accident Alarm Systems Project Position - Comply, with the following clarification: Criticality accident alarm systems are not provided in the areas of the spent fuel pools in the containment and the fuel buildings. Spent fuel is stored under water and in storage racks designed to maintain geometric spacing such that criticality is precluded. Also a criticality alarm system is not provided in the area of the new fuel storage vault. Design features and administrative controls preclude the possibility of an accidental criticality. USAR Subsections - 12.3.4, 12.5.2 CPS/USAR CHAPTER 01 1.8-80 REV. 15, JANUARY 2013 Regulatory Guide 8.13, Rev. 1 (November 1975)

Instruction Concerning Prenatal Radiation Exposure Project Position - Comply USAR Subsection - 12.5.2 Regulatory Guide 8.14, Rev. 1 (August 1977)

Personnel Neutron Dosimeters Project Position - Not applicable to Clinton Power Station. Clinton Power Station uses Dosimeters of Legal Record (DLRs) and neutron sensitive radiation measuring instruments to measure neutron exposure to personnel. DLRs are certified for accuracy and sensitivity to neutron radiation in accordance with the National Voluntary Laboratory Accreditation Program (NVLAP).

Regulatory Guide 8.15, Rev. 1 (October 1999)

Acceptable Programs for Respiratory Protection Project Position - Comply USAR Subsection - 12.5.2 Regulatory Guide 8.19, Rev. 1 (June 1979)

Occupational Dose Assessment in Light-Water Reactor Plants Design Stage Man-Rem Estimates Project Position - Comply, with the exceptions and clarifications discussed in Subsection 12.4.4. USAR Subsection - 12.4.4 Regulatory Guide 8.20, Rev. 1 (September 1979)

Applications of Bioassay For I-125 and I-131 Project Position - Not applicable at Clinton Power Station. Regulatory Guide 8.9 supersedes Regulatory Guide 8.20.

CPS/USAR CHAPTER 01 1.8-81 REV. 14, JANUARY 2011 Regulatory Guide 8.25, Rev. 1 (June 1992)

Air Sampling in the Work Place Project Position - Not applicable at Clinton Power Station.

Regulatory Guide 8.26, Rev. 0 (September 1980)

Applications of Bioassay for Fission and Activation Products Project Position - Not applicable at Clinton Power Station. Regulatory Guide 8.9 supersedes Regulatory Guide 8.26.

Regulatory Guide 8.27, Rev. 0 (March 1981)

Radiation Protection Training For Personnel at Light-Water Cooled Nuclear Power Plants.

Project Position - Comply USAR Subsection - 12.5.2, 12.5.3.5 Regulatory Guide 8.28, Rev. 0 (August 1981)

Audible Alarm Dosimeters Project Position - Comply with the following exception: The noted exception to Regulatory Guide 8.28, Paragraph C.2.c, involves the use of self performance checks for determining if electronic dosimetry is properly operating rather than use of a radiation source. USAR Subsection - 12.5.2 Regulatory Guide 8.29, Rev. 0 (July 1981)

Instructions Concerning Risks from Occupational Radiation Exposures Project Position - Comply USAR Subsection - 12.5.2, 12.5.3.5

CPS/USAR CHAPTER 01 1.9-1 REV. 11, JANUARY 2005 1.9 SYMBOLS USED IN ENGINEERING DRAWINGS The symbols used in engineering drawings are shown in Drawings M05-1000, M05-1001, 197R567, 209A4756, 209A7367, and 921D280, and in K-2999 Standards STD-EC-110 and STD-EC-111.

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CPS/USAR CHAPTER 01 1.10-1 REV. 11, JANUARY 2005 1.10 ACRONYMS The following is a list of the acronyms used in the Clinton Power Station Updated Safety Analysis Report: ABS Absolute Sum (Method) A-C(a-c) Alternating Current ACI American Concrete Institute ACRS Advisory Committee for Reactor Safeguards ADS Automatic Depressurization System AE Architect Engineer AISC American Institute of Steel Construction AISI American Iron and Steel Institute ALARA As Low As Is Reasonably Achievable ANS American Nuclear Society ANSI American National Standards Institute APED Atomic Power Equipment Department (GE)

API American Petroleum Institute API Automatic Priority Interrupt APLHGR Average Planar Linear Heat Generation Rate APRM Average Power Range Monitor ARM Area Radiation Monitor ASCE American Society of Civil Engineers ASLAB Atomic Safety & Licensing Appeals Board ASLB Atomic Safety & Licensing Board ASME American Society of Mechanical Engineers ASTM American Society for Testing Materials AT Current Transducer ATS Alarm Trip Setting (LPRM Channels)

ATWS Anticipated Transients Without Scram AWS American Welding Society B&PV Boiler & Pressure Vessel Code BA Baldwin Associates BOL Beginning of Life (fuel cycle)

BOP Balance of Plant BPWS Banked Position Withdrawal Sequence BTP Branch Technical Position (NRC) BWR Boiling Water Reactor C&I Controls & Instrumentation CCW Component Cooling Water CFR Code of Federal Regulations CGCB Containment Gas Control Boundary CGCS Combustible Gas Control System CIT Conductivity Indicator Transmitter CMAA Crane Manufacturing Association of America CMFA Common Mode Failure Analysis COC Certificate of Compliance CP Construction Permit CPM Critical Path Method CPR Critical Power Ratio CPS Clinton Power Station CPS/USAR CHAPTER 01 1.10-2 REV. 15, JANUARY 2013 CRD Control Rod Drive CRDA Control Rod Drop Accident CRDHS Control Rod Drive Hydraulic System CRO Control Room Operator CRPI Control Rod Position Indication CRT Cathode Ray Tube CRS Conductivity Recording Switch CRSI Concrete Reinforcing Steel Institute DAC Derived Air Concentration DAP Data Acquisition Processor DB Design Basis DBA Design-Basis Accident D-C (d-c) Direct Current DCP Display Control Processor DCS Display Control System (PPCS Subsystem)

DDR Deviation Disposition Request DELS Diesel Engine Lubrication System DG Diesel Generator (Diesel Engine-Generator)

DGSS Diesel Generator Starting System

DLR Dosimeter of Legal Record DNB Departure from Nucleate Boiling DOP Dioctyl Phthalate DOT U.S. Department of Transportation DPF Design Project Flood DTS Differential Temperature Switch ECA Engineering Change Authorization ECCS Emergency Core Cooling System(s)

ECN Engineering Change Notice EDS Engineering Data Systems EDT Engineering Data Transmittal EEP Engineered Equipment Procurement EFCV Excess Flow Check Valve EFDS Equipment and Floor Drainage System EHC Electrohydraulic Control EI Engineering Instructions EIC Eberline Instrument Corporation EIS Engineering Information System EOL End of Life (fuel cycle)

E/P Converter (Voltage/Pneumatic)

EP&C Engineering Production and Control EP&P Engineering Practices & Procedures ER Environmental Report ERAT Emergency Reserve Auxiliary Transformer ERDA Energy Research and Development Administration ESF Engineered Safety Feature EWA Engineering Work Authorization FA Full Arc (mode of TCV operation) FAI Fail As Is FAP Fatigue Analysis Program FC Closes on Loss of Air or Electrical Power FCD Functional Control Diagram FCF First Called For CPS/USAR CHAPTER 01 1.10-3 REV. 11, JANUARY 2005 FCR Field Change Request FCV Flow Control Valve F/D Filter Demineralizer FDDR Field Deviation Disposition Request FDI Field Disposition Instruction FHA Fuel Handling Accident FIT Flow Indicator Transmitter FLECHT Full-Length Emergency Cooling Heat Transfer FM Frequency Meter FMEA Failure Modes and Effects Analysis FO Opens on Loss of Air or Electrical Power FPCC Fuel Pool Cooling and Cleanup FPS Fire Protection System FRCS Flow Recording Controller Switch FSAR Final Safety Analysis Report GDC NRC General Design Criteria GE General Electric Company GESSAR General Electric Standard Safety Analysis Report GETAB General Electric Thermal Analysis Basis GETSCO General Electric Technical Services Company H&V Heating and Ventilating HCU Hydraulic Control Unit HEPA High-Efficiency Particulate Air/Absolute (referring to filters)

HHH High-High-High HPCS High-Pressure Core Spray HVAC Heating, Ventilating, and Air-Conditioning HX Heat Exchanger I&C Instrumentation and Control IAC Interim Acceptance Criteria (NRC) IB Inboard IBA Intermediate Break Accident ID Inside Diameter IDS Instrument Data Sheet IED Instrument Electrical Diagram IED Instrument Engineering Diagram IEEE Institute of Electrical and Electronic Engineers ILRT Integrated Leak Rate Test IO Input/Output IP or IPC Illinois Power Company IRM Intermediate Range Monitor ISI Inservice Inspection KEFF Effective Neutron Multiplication Factor LC Lock Closed LCD Local Climatological Data LCO Limiting Condition of Operation LCR Logarithm of Count Rate L/DRS Level and Density Recorder Switch LDS Leak-Detection System LFMG Low Frequency Motor Generator LFMGS Low Frequency Motor-Generator Set LHGR Linear Heat Generation Rate CPS/USAR CHAPTER 01 1.10-4 REV. 11, JANUARY 2005 LIM SW Limit Switch LIRS Level Indicator Recording Switch LO Lock Open LOCA Loss-of-Coolant Accident LOEP Loss of Electric Power LOOP Loss of Off-site Power LPAP Low Pressure Alarm Point LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray LPRM Local Power Range Monitor LPSP Low Pressure Set Point LPZ Low Population Zone LRCP Liquid Radwaste Control Panel LRS Level Recording Switch LSSS Limiting Safety System Setting MAPLHGR Maximum Average Planar Linear Heat Generation Rate MBA Misplaced Bundle Accident MCC Motor Control Center MCPR Minimum Critical Power Ratio MCR Main Control Room MDRFP Motor-Driven Reactor Feed Pump MEOD Maximum Extended Operating Domain MG Motor-Generator Set MM Modified Mercalli MOV Motor Operated Valve MREM Millirem MREM/YR Millirem Per Year MSIV Main Steam Isolation Valve MSIV-LCS Main Steam Isolation Valve Leakage Control System MSL Mean Sea Level MSL Main Steamline MTBE Mean Time Between Event MV/I Millivolt to Current Converter MVP Mechanical Vacuum Pump MWD/T Megawatt-Days of Energy Production Per Ton of UO 2 MWe Megawatts Electrical MWt Megawatts Thermal NB Nuclear Boiler NBR Nuclear Boiler Rated (power)

NC Normally Closed NCR Nonconformance Report ND Normally De-energized NDT Nondestructive Testing NDTT Nil Ductility Transition Temperature NE Normally Energized NEC National Electric Code NED Nuclear Energy Division (GE) NELPIA Nuclear Energy Liability Property Insurance Association NEPA National Environmental Policy Act NFPA National Fire Protection Association NMS Neutron-Monitoring System CPS/USAR CHAPTER 01 1.10-5 REV. 13, JANUARY 2009 NO Normally Open NOAA National Oceanic and Atmospheric Administration NPSH Net Positive Suction Head NSRB Nuclear Safety Review Board NRC U.S. Nuclear Regulatory Commission NS Nuclear System NSOA Nuclear Safety Operational Analysis NSPS Nuclear System Protection System NSSS Nuclear Steam Supply System NSSSS (NS 4) Nuclear Steam Supply System Shutoff OB Outboard OBE Operating Basis Earthquake OD Outside Diameter OFS Orificed Fuel Support OL Operating License OPRM Oscillation Power Range Monitoring ORE Occupational Radiation Exposures ORM Operational Requirements Manual OT Operational Transient OTB Onset of Transition Boiling P&ID Piping & Instrumentation Diagram PA Public Address (System) PCA Primary Coolant Activity PCIOMR Pre-Conditioning Interim Operating Management Recommendations PCMS Performance Calculation and Monitoring System PCT Peak Cladding Temperature PD Process Diagram PFD Process Flow Diagram PGCC Power Generation Control Complex PMF Probable Maximum Flood PMP Probable Maximum Precipitation PMS Performance Monitoring System (PPCS Subsystem) PORC Plant Operations Review Committee

PPCS Plant Process Computer System PPD Process Piping Diagram PQC Product Quality Certification PQL Product Quality Checklist PRM Power Range Monitor PRM Process Radiation Monitoring PRMS Process Radiation Monitoring System PRT Prompt Relief Trip PSAR Preliminary Safety Analysis Report PSTF Pressure Suppression Test Facility PSW Plant Service Water PWR Pressurized Water Reactor QA Quality Assurance QAP Quality Assurance Program QC Quality Control RAT Reserve Auxiliary Transformer RAU Remote Analog Unit RCIC Reactor Core Isolation Cooling RCIS Rod Control and Information System CPS/USAR CHAPTER 01 1.10-6 REV. 11, JANUARY 2005 RCPB Reactor Coolant Pressure Boundary RDU Remote Digital Unit RECHAR Recombination and Low Temperature Charcoal Adsorption RFP Reactor Feed Pump RFPT Reactor Feed Pump Turbine RG Regulatory Guide (NRC)

RHR Residual Heat Removal RHRS Residual Heat Removal System RM Remote Manual RMC Remote Manual Control RMCS Reactor Manual Control System RMS Remote Manual Switch RO Reactor Operator (licensed)

ROC Radwaste Operations Center RPCS Rod Pattern Control System RPIS Rod Position Information System RPM Radiation Protection Manager RPS Reactor Protection System RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup RWE Rod Withdrawal Error RWL Rod Withdrawal Limiter S&L Sargent & Lundy SACF Single Active Component Failure SAR Safety Analysis Report SBA Small Break Accident SDS System Design Specifications SEF Single Equipment Failure SER Safety Evaluation Report SGTS Standby Gas Treatment System SGTSET Standby Gas Treatment System Equipment Train SIP Standby Information Panel SIT Structural Integrity Test SJAE Steam Jet Air Ejector SLC Standby Liquid Control SOE Single Operator Error SOF Single Operator Failure SPCU Suppression Pool Cleanup SPF Standard Project Flood SPMU Suppression Pool Makeup SPS Standard Project Storm SQRTOR Square Root Convertor SRM Source Range Monitor SRO Senior Reactor Operator (licensed)

SRP Standard Review Plan (NRC) SRSS Square Root of the Sum of Squares SRV Safety/Relief Valve SS Safe Shutdown SS Selector Switch SSA Selective Switch Automatic CPS/USAR CHAPTER 01 1.10-7 REV. 15, JANUARY 2013 SSE Safe Shutdown Earthquake SSW Shutdown Service Water SW Service Water SWS Service Water System SWU Separative Work Unit TBCCW Turbine Building Closed Cooling Water TBCCWS Turbine Building Closed Cooling Water System TCV Turbine Control Valve TDRFP Turbine-Driven Reactor Feed Pump TDS Total Dissolved Solids TDS Time Delay Switch TG Turbine-Generator

TIP Traversing Incore Probe TPM Thermal Power Monitor TQOS Torque Overload Switch TQRS Torque Recorder Switch TRS Temperature Recorder Switch TRU Test and Reconfigure Unit TSI Turbine Supervisor Instruments TSV Turbine Stop Valve UBC Uniform Building Code UHS Ultimate Heat Sink URC Ultrasonic Resin Cleaner USAR Updated Safety Analysis Report USGS United States Geological Survey UT Ultrasonic Testing Vac Volts-Alternating Current VARM VAR Meter VART VAR Transducer VBWR Vallecitos Boiling Water Reactor Vdc Volts-Direct Current VM Voltmeter VRF Velocity Range Factor VWO Valves Wide Open WHM Watt-hour meter WT Watt Transducer XFMR Transformer