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Nuclear Engineering : an Introduction

Author: K Almenas; R Lee
Publisher: Berlin, Heidelberg : Springer Berlin Heidelberg, 1992.
Edition/Format:   eBook : Document : EnglishView all editions and formats
Summary:
***VERKAUFSKATEGORIE*** 1 e This textbook covers the core subjects of nuclear engineering. Developed to meet the needs of today's students and nuclear power plant operators, the text establishes a framework for the various areas of knowledge that comprise the field and explains rather than just defines the relevant physical phenomena. For today's engineer the principal analytical design tool is the personal  Read more...
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Genre/Form: Electronic books
Additional Physical Format: Print version:
Material Type: Document, Internet resource
Document Type: Internet Resource, Computer File
All Authors / Contributors: K Almenas; R Lee
ISBN: 9783642488764 3642488765
OCLC Number: 851829454
Description: 1 online resource (x, 566 pages 191 illustrations)
Contents: 1 Introduction.- 1.1 The General and the Unique.- 1.2 The Law of Laws.- 1.3 A Neutron Population Balance.- 2 Neutron/Nuclei Balance - The Fission Source.- 2.1 Fission Process as an Example.- 2.1.1 Nuclear Binding Energy.- 2.1.2 Excited States or Energy Levels in Atoms and Nuclei.- 2.1.3 The Compound Nucleus.- 2.1.4 The Fission Event.- 2.1.5 Fissile, Fissionable, and Fertile Isotopes.- 2.2 Radioactive Decay.- 2.2.1 Number (Atomic/Molecular) Density, N.- 2.2.2 Statistical and Quantized Nature of Radioactivity.- 2.2.3 The General Radioactive Isotope Balance.- 2.2.4 Decay Constant, Half-Life, and Mean Life.- 2.2.5 Fission Products in Nuclear Reactor Cores.- Rroblems.- 3 Neutron Interaction with Matter.- 3.1 Macroscopic and Microscopic Cross Sections.- 3.2 Neutron Fluxes, Currents, and Beams.- 3.3 Cross-Sections Type.- 3.4 Macroscopic Cross-Sections.- 3.5 Energy Dependence of Cross-Sections.- 3.6 The "Six Factor" Formula.- 3.7 The SIXFAC Program.- 3.8 Calculation of k?.- 3.9 Calculation of k? for a "fast reactor".- 3.10 Effective multiplication factor, keff.- Problems.- 4 Neutron Diffusion - Basic Concepts.- 4.1 Fick's Law.- 4.2 The Diffusion Coefficient.- 4.3 Fick's Law Analogues in Other Branches of Science.- 4.4 The One-group Diffusion Equation.- 4.5 Boundary Conditions.- 4.6 Solution to the One-group Diffusion Equation.- 4.6.1 Solution for a Plane Source.- 4.6.2 Solution for a Point Source.- Problems.- 5 Neutron Balance - Energy.- 5.1 Neutron Moderation.- 5.1.1 The Importance of the Scattering Interaction.- 5.1.2 Evaluation of the Neutron Energy After Scattering.- 5.1.3 The COLLIDE and SCATEREL Programs.- 5.2 Neutron Energy Distribution in the Epithenmal Region.- 5.3 Properties and Classification of Moderators.- 5.4 Thermal Neutrons.- 5.4.1 The Maxwellian Distribution.- 5.4.2 The Thermal Flux.- 5.4.3 Non-1/V Factors.- 5.4.4 Mathematical Properties of a Maxwellian Distribution.- Problems.- 6 Criticality.- 6.1 Criticality and the Neutron Balance Equation.- 6.2 Basic Relationships. The `Text Book' Slab Core.- 6.3 Criticality Condition for the Generic Bare Core.- 6.4 Criticality for Multienergy Group Neutron Balances.- 6.5 Finite Difference Solution Methods.- 6.6 The MULTIDIF (MULTI-group DIF-usion) Code.- Problems.- 7 Neutron Balance - Time.- 7.1 Prompt and Delayed Neutrons.- 7.1.1 Steady State and Time Dependent Neutron Balances.- 7.1.2 Characteristics of Delayed Neutrons.- 7.1.3 Neutron Generation Time.- 7.1.4 Prompt and Delayed Criticality.- 7.1.5 Definition of Reactivity Units.- 7.2 Solution of Kinetics Equations.- 7.2.1 Basic Assumptions Made.- 7.2.2 Single Neutron Group Kinetics.- 7.2.3 Kinetic Equation with Delayed Neutrons.- 7.2.4 The Asymptotic Period.- 7.2.5 The `Prompt' Jump.- 7.2.6 Estimation of Small Reactivities.- 7.3 Reactor Control Methods.- 7.3.1 Types of Reactivity Control.- 7.3.2 PWR Pin Type Control Rods.- 7.3.3 BWR (Cruciform) Control Rods.- 7.3.4 Centrally Located Control Rod.- 7.4 Control Practice.- 7.4.1 Control Rod Worth Curves.- 7.4.2 Impact of a Control-Rod on the Neutron Flux.- 7.4.3 Soluble Poison (Chemical Shim) Control.- 7.5 Temperature and Power Coefficients of Reactivity.- 7.5.1 Reactivity Coefficients.- 7.5.2 The Fuel Temperature (Doppler) Reactivity Coefficient.- 7.5.3 The Moderator Reactivity Coefficient.- 7.5.4 The `Void' Coefficient of Reactivity.- 7.5.5 The Power Coefficient of Reactivity.- 7.5.6 The SIXFACT Code.- 7.6 Fission Product Effects.- 7.6.1 Fission Product Types.- 7.6.2 High Cross Section Fission Products: Xe-135.- 7.6.3 Xe Transients.- 7.6.4 Sm-149 Poisoning.- 7.6.5 Fuel Depletion.- 7.6.6 Long Term Fission Product Buildup.- Problems.- 8 Gamma and Neutron Radiation Effects.- 8.1 Importance of Gamma Rays.- 8.1.1 Fundamental Gamma Ray-Matter Interaction Modes.- 8.1.2 Attenuation Coefficients.- 8.1.3 Energy Deposition.- 8.2 Radiation Units.- 8.2.1 From Source to Dose.- 8.2.2 Units of Source Intensity. Activity.- 8.2.3 Units of the Radiation Field. Exposure.- 8.2.4 Units of Energy Deposition. Dose.- 8.3 Radiation Sources.- 8.3.1 Natural Radiation Sources.- 8.3.2 Manmade Radiation Sources.- 8.3.3 Effects of Radon.- 8.4 Biological Effects of Radiation.- 8.4.1 Radiation Damage Mechanisms.- 8.4.2 Relative Biological Effectiveness.- 8.4.3 Stochastic and Nonstochastic Effects.- 8.4.4 Acute, Latent and Genetic Radiation Effects.- 8.4.5 Calculations. Effective Gamma Ray Doses.- 8.4.6 Calculation External Neutron Doses.- 8.5 Radiation Protection Standards.- 8.5.1 Historical Overview.- 8.5.2 Current Standards. External Radiation.- 8.5.3 Internal Radiations Sources. Calculations.- 8.5.4 Safeguard Standards for Internal Radiation.- Problems.- 9 Shielding.- 9.1 Basic Concepts.- 9.1.1 Characteristics of Shielding Problems.- 9.1.2 Spread of Radiation. Point Source.- 9.1.3 Spread of Radiation. Sources Having Simple Geometries.- 9.2 Buildup Factors.- 9.2.1 Uncolided and Scattered Radiation Beam Components.- 9.2.2 Definition of Buildup Factors.- 9.2.3 Example of the Use of Buildup Factors. Plane Source.- 9.3 Basic Shield Geometries.- 9.3.1 Shielded Infinite Plane Source.- 9.3.2 The Line Source.- 9.3.3 Volumetric Radiation Sources.- 9.4 Computer Methods in Gamma Shield Design.- 9.4.1 Generalized Numerical Integration Methods (MathCAD).- 9.4.2 Point Kernel Methods.- 9.5 Reactor Shielding Problems.- 9.5.1 Fission from the Shielding Perspective.- 9.5.2 Overview of Radiation Types in an Operating Reactor.- 9.5.3 Energy Dependence of Core Radiation.- 9.5.4 Neutron Shielding. The `Removal' Cross Section.- 9.5.5 Methods for Estimating Effectiveness of Core Shields.- 9.5.6 Activation Induced Radiation.- 9.5.7 Coolant Activation.- 9.6 Neutron Shielding. Exact Methods.- 9.6.1 Transport Theory. Basic Definitions.- 9.6.2 Transport Theory. Shielding Applications.- 9.6.3 Coupled Neutron-Gamma Cross Sections.- 9.6.4 Monte Carlo Theory.- Problems.- 10 Core Heat Removal.- 10.1 Core Energy Balance.- 10.1.1 Overview.- 10.1.2 General Energy Balance.- 10.1.3 Energy Balance of PWR's.- 10.1.4 Energy Balance for BWR Cores.- 10.2 Energy Sources.- 10.2.1 Overview of Fission Energy.- 10.2.2 Fuel Heat Sources.- 10.2.3 Decay Heat.- 10.3 The Heat Transfer Path.- 10.3.1 Heat Conduction Equation.- 10.3.2 Temperature Distribution Calculations for a Fuel Rod.- 10.3.3 Effect of Temperature Dependent Conductivity.- 10.3.4 Convective Heat Transfer Coefficients.- 10.3.5 The FUELROD Program.- 10.4 DNB and CHF Ratios.- 10.4.1 Boiling Correlations.- 10.4.2 CHF and DNB Calculations.- 10.4.3 Hot Channel Factors.- Problems.- 11 Reactor Licensing.- 11.1 The Nuclear Regulatory Commission: Historical Background.- 11.2 The NRC: Organizational Structure and the "Licensing Process".- 11.2.1 Organizational Structure.- 11.2.2 The "Licensing Process".- 11.3 Title 10 of the Code of Federal Regulations (10 CFR).- 11.4 The Design-Basis Accidents.- 11.4.1 Can Accidents be Designed?.- 11.4.2 The Design Base (DB) Accidents.- 11.4.3 The "Small Break" Loss of Coolant Accident (SBLOCA).- 11.5 The Multiple Barriers.- 11.5.1 The Fuel.- 11.5.2 The Cladding.- 11.5.3 The Primary Coolant.- 11.5.4 Reactor Vessel.- 11.5.5 The Containment Building.- 11.5.6 Large Dry Containment.- 11.5.7 Pressure Suppression Containments.- 11.6 Fission Product Release.- 11.6.1 The "Source" Term.- 11.6.2 Buildup of Fission Products Inventory.- 11.6.3 Leakage from Buildings.- 11.7 Atmospheric Dispersion.- 11.7.1 Meteorology of Atmospheric Dispersion.- 11.7.2 Diffusion Relationship.- 11.8 Class Nine Accidents.- 11.8.1 Characteristics of a Class 9 Accident.- 11.8.2 The Class 9 Accident Scenario.- 11.8.3 Generation of Hydrogen.- 11.9 Accident Risk Analysis.- 11.9.1 Society and Risk.- 11.9.2 Quantification of Risk.- Problems.
Responsibility: by K. Almenas, R. Lee.

Abstract:

Developed to meet the needs of today's students and nuclear power plant operators, the text establishes a framework for the various areas of knowledge that comprise the field and explains rather than  Read more...

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