Nuclear Systems Volume I - Todreas Neil E.; Kazimi Mujid S | Libro + Cd-Rom Taylor & Francis 09/2011 -

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Nuclear Systems Volume I Thermal Hydraulic Fundamentals, Second Edition


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Lingua: Inglese
Pubblicazione: 09/2011
Edizione: Edizione nuova, 2° edizione

Note Editore

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. Tables for Computation available for download at 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.


Principal Characteristics of Power ReactorsIntroductionPower CyclesPrimary Coolant SystemsReactor CoresFuel AssembliesAdvanced Water- and Gas-Cooled Reactors (Generation III And III+)Advanced Thermal and Fast Neutron Spectrum Reactors (Generation IV)ReferencesProblemsThermal Design Principles and ApplicationIntroductionOverall Plant Characteristics Influenced by Thermal Hydraulic ConsiderationsEnergy Production and Transfer ParametersThermal Design LimitsThermal Design MarginFigures of Merit for Core Thermal PerformanceThe Inverted Fuel ArrayThe Equivalent Annulus ApproximationReferencesProblemsReactor Energy DistributionIntroductionEnergy Generation and DepositionFission Power and Calorimetric (Core Thermal) PowerPower Profiles in Reactor CoresEnergy Generation Rate within a Fuel PinEnergy Deposition Rate within The ModeratorShutdown Energy Generation RateStored Energy SourcesReferencesProblemsTransport Equations for Single-Phase FlowIntroductionMathematical RelationsLumped Parameter Integral ApproachDistributed Parameter Integral ApproachDifferential Conservation EquationsTurbulent FlowReferencesProblemsTransport Equations for Two-Phase FlowIntroductionAveraging Operators for Two-Phase FlowVolume-Averaged PropertiesArea-Averaged PropertiesMixture Equations for One-Dimensional FlowControl-Volume Integral Transport EquationsOne-Dimensional Space-Averaged Transport EquationsReferencesProblems Thermodynamics of Nuclear Energy Conversion Systems: Nonflow and Steady Flow: First and Second Law ApplicationsIntroductionNonflow ProcessThermodynamic Analysis of Nuclear Power PlantsThermodynamic Analysis of a Simplified Pwr SystemMore Complex Rankine Cycles: Superheat, Reheat, Regeneration, and Moisture SeparationSimple Brayton CycleMore Complex Brayton CyclesReferenceProblemsThermodynamics of Nuclear Energy Conversion Systems: Nonsteady Flow First Law AnalysisIntroductionContainment Pressurization ProcessResponse of a PWR Pressurizer to Load ChangesReferencesProblems Thermal Analysis of Fuel ElementsIntroductionHeat Conduction in Fuel ElementsThermal Properties of UO2 and MOxTemperature Distribution in Plate Fuel ElementsTemperature Distribution in Cylindrical Fuel PinsTemperature Distribution in Restructured Fuel ElementsThermal Resistance Between Fuel and CoolantReferencesProblemsSingle-Phase Fluid MechanicsApproach to Simplified Flow AnalysisInviscid FlowViscous FlowLaminar Flow Inside a ChannelTurbulent Flow Inside a ChannelPressure Drop in Rod BundlesReferencesProblemsSingle-Phase Heat TransferFundamentals of Heat Transfer AnalysisLaminar Heat Transfer in a PipeTurbulent Heat Transfer: Mixing Length ApproachTurbulent Heat Transfer: Differential ApproachHeat Transfer Correlations in Turbulent FlowReferencesProblemsTwo-Phase Flow DynamicsIntroductionFlow RegimesFlow ModelsOverview of Void Fraction and Pressure Loss CorrelationsVoid Fraction CorrelationsPressure-Drop RelationsCritical FlowReferencesProblemsPool BoilingIntroductionNucleationThe Pool Boiling CurveHeat Transfer RegimesSurface Effects in Pool BoilingCondensation Heat TransferReferencesProblemsFlow BoilingIntroductionHeat Transfer Regions and Void Fraction/Quality DevelopmentHeat Transfer Coefficient CorrelationsCritical Condition or Boiling CrisisReferencesProblemsSingle Heated Channel: Steady-State AnalysisIntroductionFormulation of One-Dimensional Flow EquationsDelineation of Behavior ModesThe Lwr Cases Analyzed in Subsequent SectionsSteady-State Single-Phase Flow in a Heated ChannelSteady-State Two-Phase Flow in a Heated Channel Under Fully Equilibrium (Thermal and Mechanical) ConditionsSteady-State Two-Phase Flow in a Heated Channel Under Nonequilbrium ConditionsReferencesProblemsAPPENDICESAppendix A: NOMENCLATUREAppendix B: PHYSICAL AND MATHEMATICAL CONSTANTSAppendix C: UNIT SYSTEMSAppendix D: MATHEMATICAL TABLESAppendix E: THERMODYNAMIC PROPERTIESAppendix F: THERMOPHYSICAL PROPERTIES OF SOME SUBSTANCESAppendix G: DIMENSIONLESS GROUPS OF FLUID MECHANICS AND HEAT TRANSFERAppendix H: MULTIPLYING PREFIXESAppendix I: LIST OF ELEMENTSAppendix J: SQUARE AND HEXAGONAL ARRAY DIMENSIONSAppendix K PARAMETERS FOR TYPICAL PWR AND BWR-5 REACTORS


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.

Altre Informazioni



Condizione: Nuovo
Dimensioni: 9.25 x 6.125 in Ø 3.25 lb
Formato: Copertina rigida
Illustration Notes:343 b/w images, 134 tables and 1303+ equations-SEE NOTES 6th & 1st print-10/15-whole book re-done
Pagine Arabe: 1034

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