Nuclear Engineering
An Introduction
(Sprache: Englisch)
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...
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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 computer. The text takes advantage of this recent development. PC programs are provided which either expand the computational range accessible to the student, or serve to illustrate the relevant physical phenomena. Some of the included programs are simplified versions of computational procedures used in the field and can be used as training tool for design calculations. The text devotes special attention to subjects which have an impact on the safe operation of nuclear power reactors. This includes the design of safety optimized core configurations, the physical mechanisms underlying the various reactivity coefficients, and the calibration procedures for control rods. A final chapter is devoted to the licensing and safety evaluation of power reactors.
Inhaltsverzeichnis zu „Nuclear Engineering “
1 Introduction1.1 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
... mehr
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
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
... weniger
Bibliographische Angaben
- Autoren: K. Almenas , R. Lee
- 2014, Softcover reprint of the original 1st ed. 1992., 566 Seiten, Maße: 15,9 x 23,8 cm, Kartoniert (TB), Englisch
- Verlag: Springer
- ISBN-10: 3642488781
- ISBN-13: 9783642488788
- Erscheinungsdatum: 13.04.2014
Sprache:
Englisch
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