Nuclear Systems Volume I: Thermal Hydraulic Fundamentals, Second Edition

Nuclear power is in the midst of a generational change--with new reactor designs, plant subsystems, fuel concepts, and other information that must be explained and explored--and after the 2011 Japan disaster, nuclear reactor technologies are, of course, front and center in the public eye. Written by leading experts from MIT, Nuclear Systems Volume I: Thermal Hydraulic Fundamentals, Second Edition provides an in-depth introduction to nuclear power, with a focus on thermal hydraulic design and analysis of the nuclear core. A close examination of new developments in nuclear systems, this book will help readers--particularly students--to develop the knowledge and design skills required to improve the next generation of nuclear reactors. Includes a CD-ROM with Extensive Tables for Computation Intended for experts and senior undergraduate/early-stage graduate students, the material addresses: Different types of reactors Core and plant performance measures Fission energy generation and deposition Conservation equations Thermodynamics Fluid flow Heat transfer Imparting a wealth of knowledge, including their longtime experience with the safety aspects of nuclear installations, authors Todreas and Kazimi stress the integration of fluid flow and heat transfer, various reactor types, and energy source distribution. They cover recent nuclear reactor concepts and systems, including Generation III+ and IV reactors, as well as new power cycles. The book features new chapter problems and examples using concept parameters, and a solutions manual is available with qualifying course adoption.

Table of Contents:
Principal Characteristics of Power Reactors Introduction Power Cycles Primary Coolant Systems Reactor Cores Fuel Assemblies Advanced Water- and Gas-Cooled Reactors (Generation III And III+) Advanced Thermal and Fast Neutron Spectrum Reactors (Generation IV) References Problems Thermal Design Principles and Application Introduction Overall Plant Characteristics Influenced by Thermal Hydraulic Considerations Energy Production and Transfer Parameters Thermal Design Limits Thermal Design Margin Figures of Merit for Core Thermal Performance The Inverted Fuel Array The Equivalent Annulus Approximation References Problems Reactor Energy Distribution Introduction Energy Generation and Deposition Fission Power and Calorimetric (Core Thermal) Power Power Profiles in Reactor Cores Energy Generation Rate within a Fuel Pin Energy Deposition Rate within The Moderator Shutdown Energy Generation Rate Stored Energy Sources References Problems Transport Equations for Single-Phase Flow Introduction Mathematical Relations Lumped Parameter Integral Approach Distributed Parameter Integral Approach Differential Conservation Equations Turbulent Flow References Problems Transport Equations for Two-Phase Flow Introduction Averaging Operators for Two-Phase Flow Volume-Averaged Properties Area-Averaged Properties Mixture Equations for One-Dimensional Flow Control-Volume Integral Transport Equations One-Dimensional Space-Averaged Transport Equations References Problems Thermodynamics of Nuclear Energy Conversion Systems: Nonflow and Steady Flow: First and Second Law Applications Introduction Nonflow Process Thermodynamic Analysis of Nuclear Power Plants Thermodynamic Analysis of a Simplified Pwr System More Complex Rankine Cycles: Superheat, Reheat, Regeneration, and Moisture Separation Simple Brayton Cycle More Complex Brayton Cycles Reference Problems Thermodynamics of Nuclear Energy Conversion Systems: Nonsteady Flow First Law Analysis Introduction Containment Pressurization Process Response of a PWR Pressurizer to Load Changes References Problems Thermal Analysis of Fuel Elements Introduction Heat Conduction in Fuel Elements Thermal Properties of UO2 and MOx Temperature Distribution in Plate Fuel Elements Temperature Distribution in Cylindrical Fuel Pins Temperature Distribution in Restructured Fuel Elements Thermal Resistance Between Fuel and Coolant References Problems Single-Phase Fluid Mechanics Approach to Simplified Flow Analysis Inviscid Flow Viscous Flow Laminar Flow Inside a Channel Turbulent Flow Inside a Channel Pressure Drop in Rod Bundles References Problems Single-Phase Heat Transfer Fundamentals of Heat Transfer Analysis Laminar Heat Transfer in a Pipe Turbulent Heat Transfer: Mixing Length Approach Turbulent Heat Transfer: Differential Approach Heat Transfer Correlations in Turbulent Flow References Problems Two-Phase Flow Dynamics Introduction Flow Regimes Flow Models Overview of Void Fraction and Pressure Loss Correlations Void Fraction Correlations Pressure-Drop Relations Critical Flow References Problems Pool Boiling Introduction Nucleation The Pool Boiling Curve Heat Transfer Regimes Surface Effects in Pool Boiling Condensation Heat Transfer References Problems Flow Boiling Introduction Heat Transfer Regions and Void Fraction/Quality Development Heat Transfer Coefficient Correlations Critical Condition or Boiling Crisis References Problems Single Heated Channel: Steady-State Analysis Introduction Formulation of One-Dimensional Flow Equations Delineation of Behavior Modes The Lwr Cases Analyzed in Subsequent Sections Steady-State Single-Phase Flow in a Heated Channel Steady-State Two-Phase Flow in a Heated Channel Under Fully Equilibrium (Thermal and Mechanical) Conditions Steady-State Two-Phase Flow in a Heated Channel Under Nonequilbrium Conditions References Problems APPENDICES Appendix A: NOMENCLATURE Appendix B: PHYSICAL AND MATHEMATICAL CONSTANTS Appendix C: UNIT SYSTEMS Appendix D: MATHEMATICAL TABLES Appendix E: THERMODYNAMIC PROPERTIES Appendix F: THERMOPHYSICAL PROPERTIES OF SOME SUBSTANCES Appendix G: DIMENSIONLESS GROUPS OF FLUID MECHANICS AND HEAT TRANSFER Appendix H: MULTIPLYING PREFIXES Appendix I: LIST OF ELEMENTS Appendix J: SQUARE AND HEXAGONAL ARRAY DIMENSIONS Appendix K PARAMETERS FOR TYPICAL PWR AND BWR-5 REACTORS

About the Author :
Dr. Neil Todreas is professor emeritus at MIT. He has extensive nuclear power experience, having led an industry review group on the Three Mile Island situation from 1983-1988 and served on the NRC's Reactor Safety Research Committee. In addition to his part-time teaching and research, Dr. Todreas continues to be a leading consultant to industry and government. He is a Fellow at the ASME and a member of the national academy of engineering. Dr. Mujid Kazimi is a professor and former head of the Department of Nuclear Engineering at MIT. He also has extensive nuclear power experience, having served on the Board of Managers of the Idaho National Energy Laboratory. He is also a Fellow at the American Nuclear Society and the AAAS, and a member of the AIChE, ASME and ASEE. Dr. Kazimi has been involved with several nuclear safety studies throughout his career, covering reactor systems, as well as their fuel cycles.