Nuclear Reactor Physics and Engineering E-Book

John C. Lee

• Herausgeber: John Wiley & Sons
• Kategorie: Wissenschaft und neue Technologien
• Sprache: Englisch

Seitenzahl: 1027

Veröffentlichungsjahr: 2024

Beschreibung

Essential guide to analyzing nuclear energy systems, with focus on reactor physics, fuel cycle, system dynamics, thermal-hydraulics, and economics.

Nuclear Reactor Physics and Engineering highlights efforts in utilizing low enrichment uranium fuel as a substitute for carbon-based fuels in energy generation and provides an overview of important aspects of nuclear reactor physics utilizing the neutron diffusion equation for major reactor designs and MATLAB software for system analysis, with exercises illustrating key points and design parameters as supplementary material.

This revised and updated Second Edition reflects key findings of the 2023 National Academy of Sciences (NAS) report and discusses physical and engineering characteristics of advanced nuclear reactors, especially in the form of small modular reactors that have the potential to provide enhanced safety and economics, as well as effective long-term management of used nuclear fuel in geological repositories.

Key topics explored in the updated edition of Nuclear Reactor Physics and Engineering include:

• Impact of the use of high-assay low enrichment uranium (HALEU) fuel as a new efficient nuclear fuel
• Advantages resulting from combined uses of light water reactor and sodium-cooled fast reactor with fuel reprocessing
• Fundamental nuclear reactor physics, nuclear reactor system analysis, and lattice physics analysis for reactor cores
• Nuclear fuel cycle analysis, nuclear plant simulation and control, and management of used nuclear fuel
• Economic analysis of nuclear electricity and thermal-hydraulic analysis of nuclear systems.

With a wealth of all-new information detailing the state of the art in the field, Nuclear Reactor Physics and Engineering is an invaluable reference on the subject for undergraduate and graduate students in nuclear engineering, as well as practicing engineers involved with nuclear power plants.

Leseprobe

Nuclear Reactor Physics and Engineering

John C. Lee

University of Michigan

Ann Arbor, Michigan

Second Edition

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List of Tables

Table 1.1 Key features of three Generation IV plants.

Table 1.2 Key attributes of advanced nuclear reactors under development.

Table 2.1 Fission energy breakdown for U, U, U, Pu, and Pu.

Table 2.2 Delayed neutron data for , , , , and for incident neutron energy of 0.1 eV, ENDF/B-VIII.0 MF = 1, 5, 7, 9, 11, MT = 455.

Table 2.3 Contents of evaluated nuclear data file.

Table 5.1 Thermal diffusion properties of moderators.

Table 5.2 Flux and geometric buckling for bare critical reactors.

Table 5.3 Number densities and microscopic cross sections for a PWR core.

Table 5.4 Macroscopic cross sections and absorption rates.

Table 11.1 Extrapolation distance in units of transport mean free path for a black cylinder.

Table 11.2 Comparison of thermal-group lattice parameters calculated in Examples 11.1 through 11.3.

Table 12.1 Comparison of parameter for key fissile nuclides.

Table 12.2 Evolution of a three-batch core beginning with a uniform first cycle with equilibrium cycle length .

Table 12.3 Evolution of a three-batch core beginning with a non-uniform first cycle approaching equilibrium cycle length .

Table 12.4 Components of excess reactivity () over cycle for LWRs.

Table 12.5 Spectrum-average cross sections for thermal and fast-spectrum reactors.

Table 12.6 Used nuclear fuel radioactivities at 100 yr (kCi/GWe-yr).

Table 12.7 Used nuclear fuel radioactivities at yr (Ci/GWe-yr).

Table 12.8 Ingestion radiotoxicity of UNF at yr (MSv/GWe-yr).

Table 12.9 UNF inventory and repository metrics of advanced nuclear reactors.

Table 13.1 Equations of mass, momentum, and energy conservation.

Table 13.2 Generalized form of fluid transport equations.

Table 13.3 Thermal-hydraulic parameters of AP600 and SBWR designs.

Table 14.1 Representative feedback coefficients and temperature rises.

Table 15.1 Overnight capital cost for 1.0 GWe LWR plant.

Table 15.2 Cost estimates for nuclear fuel cycle.

Table 15.3 Cost estimates for various energy technologies.

Table 15.4 Effective levelized cost of electricity.

Table 15.5 Energy return on investment for electrical energy technologies.

Table 16.1 Stability index and oscillation period measurements at the RGE Plant.

Table 17.1 Escape probability as a function of mean chord length.

Table A.1 Key physical constants.

Table B.1 Key features of major reactor types.

Table C.1 Legendre polynomials.

Table C.2 Zeros of the Bessel functions of the first kind.

Table D.1 A short table of Laplace transforms.

Table D.2 A short table of Fourier transforms.

Table E.1 Phase plane solution with constraints for Example E.1.

Table E.2 Phase plane solution for Example E.2.

List of Figures

Figure 1.1 Nuclear power plant evolution.

Figure 1.2 Overall layout of a PWR plant.

Figure 1.3 Schematic diagram of a PWR plant.

Figure 1.4 Cutaway view of a PWR pressure vessel illustrating key components including fuel elements and supporting structures.

Figure 1.5 Core and fuel assembly structure of a typical PWR plant. (a) Top view of the reactor core, comprising fuel assemblies and other structures inside the reactor vessel and (b) sketch of a fuel assembly illustrating fuel rods, spacer grids, rod cluster control elements, and other components.

Figure 1.6 Cross-section view of PWR fuel assemblies for the AP1000 design.

Figure 1.7 Schematic diagram of a BWR plant. Abbreviations: BPV = bypass valve, CV = control valve, CBP = condensate booster pump. CP = condensate pump, F/D = filter demineralizer, HTX = heat exchanger, SRV = safety relief valve, SV = stop valve.

Figure 1.8 Cutaway view of Mark I BWR containment structure.

Figure 1.9 Cutaway view of a BWR pressure vessel illustrating detailed coolant flow and core spray arrangement.

Figure 1.10 BWR fuel bundle cluster illustrating the W-W and N-N gaps.

Figure 1.11 Schematic diagram of the ESBWR safety system. Abbreviations: DPV = depressurization valve, IC = isolation condenser, SRV = safety relief valve, GDCS = gravity-driven cooling system, PCC = passive containment cooling.

Figure 1.12 Sodium-cooled fast reactor plant.

Figure 1.13 Top view of the reactor core of a SFR plant.

Figure 1.14 Very high-temperature reactor plant.

Figure 1.15 Top view inside the reactor vessel of a VHTR plant.

Figure 1.16 TRISO particle, pin cell, and prismatic fuel assembly for the VHTR plant.

Figure 1.17 Molten-salt reactor plant.

Figure 1.18 Top view inside a MSR reactor vessel.

Figure 1.19 Schematic illustration of the NuScale module.

Figure 1.20 Micro modular reactor.

Figure 2.1 A collimated beam of neutron incident on a slab.

Figure 2.2 Energy levels for a compound nucleus.

Figure 2.3 Fraction of fission product released vs. mass number for thermal fission of U for incident neutron energy of 0.1 eV, ENDF/B-VIII.0 MT = 454, MF = 8.

Figure 2.4 Number of total fission neutrons emitted per fission for U and Pu, ENDF/B-VIII.0 MF = 1, MT = 452.

Figure 2.5 Number of delayed neutrons released as a function of incident neutron energy for U and Pu, ENDF/B-VIII.0 MF = 1, MT = 455.

Figure 2.6 Energy spectrum of fission neutrons emitted for U, ENDF/B-VIII.0 MF = 5, MT = 18 and 455.

Figure 2.7 Velocities before and after the collision in Lab and CM systems.

Figure 2.8 Relationship between velocities and scattering angles (a) before and (b) after the collision.

Figure 2.9 Resonance cross section as a function of neutron energy.

Figure 2.10 Scattering of collimated beam into solid angle around in distance .

Figure 2.11 Solid angle with azimuthal symmetry.

Figure 2.12 Elastic scattering kernel as a function of outgoing neutron energy.

Figure 2.13 Total cross section of B, ENDF/B-VIII.

Figure 2.14 Total cross section of C, ENDF/B-VIII.

Figure 2.15 Radiative capture cross section for U, ENDF/B-VIII.

Figure 2.16 Radiative capture cross section for Pu, ENDF/B-VIII.

Figure 3.1 Differential volume elements in physical and velocity spaces.

Figure 3.2 Differential cylinder for visualizing angular flux as neutron current.

Figure 3.3 Neutron flux for a collimated neutron beam.

Figure 3.4 Projection of a unit cross-sectional area.

Figure 3.5 Relationship between vector current and net current .

Figure 3.6 Maxwell–Boltzmann distribution as a function of speed.

Figure 4.1 Unit cross-sectional area for the negative partial current.

Figure 4.2 Directional vector projected via polar and azimuthal angles.

Figure 4.3 Physical interpretation of transport mean free path .

Figure 5.1 Linear extrapolation of flux at a free surface.

Figure 5.2 Two material regions separated by vacuum.

Figure 5.3 Planar source in slab geometry.

Figure 5.4 Closed contour in the Fourier domain -space.

Figure 5.5 Extrapolated boundary for slab geometry.

Figure 5.6 Two-region slab with plane source.

Figure 5.7 Construction of a plane source from annular ring sources.

Figure 5.8 Three lowest-order flux modes for slab reactor of height .

Figure 5.9 Life cycle of neutrons illustrating the six-factor formula.

Figure 6.1 Discretization scheme for the flux and flux derivative.

Figure 6.2 Discretized flux distribution at the vacuum boundary.

Figure 6.3 Flux distribution near a reflecting boundary at .

Figure 6.4 Inner and outer iterations for solution of the diffusion equation.

Figure 6.5 Finite-difference mesh structure for .

Figure 6.6 2-D discretization scheme.

Figure 6.7 Two-dimensional finite-difference structure.

Figure 6.8 2-D line relaxation scheme.

Figure 6.9 Successive relaxation scheme.

Figure 6.10 NEM geometry for solution in the -direction.

Figure 6.11 Illustration of the homogenous and heterogeneous fluxes.

Figure 6.12 Arnoldi’s procedure converting square matrix to upper Hessenberg matrix , which is generally smaller than in size.

Figure 7.1 Lethargy variable and energy group structure.

Figure 7.2 Comparison of one-group and two-group flux distributions for a reflected slab reactor: (a) One-group flux distribution showing a monotonic decrease across the core-reflector interface (b) two-group flux distributions, indicating thermal flux peaking in the reflector.

Figure 8.1 Schematics of decay chains for fission product Br.

Figure 8.2 Transient and stable solutions of the point kinetic equations with dollar.

Figure 8.3 Reactivity versus roots of the inhour equation

Figure 8.4 Transfer function representing the output-to-input ratio.

Figure 8.5 Open-loop transfer function connecting control input to output together with controller .

Figure 8.6 Simulink setup for the pulse source solution of Example 8.2.

Figure 8.7 Simulink pulse source output of Example 8.2.

Figure 8.8 Phase plane solution of the Ergen–Weinberg model.

Figure 8.9 Time-domain behavior of the Ergen–Weinberg model.

Figure 8.10 Time-domain behavior of the Nordheim–Fuchs power excursion model.

Figure 8.11 Inverse multiplication as a function of fuel mass.

Figure 8.12 Schematic diagram of reactor transfer function with thermal-hydraulic feedback function .

Figure 8.13 Generation of Nyquist diagram via traversing contour in the right-hand half of the -plane. Contour in diagram (a) is extended with the radius , while the phase margin (PM) and gain margin (GM) in diagram (b) indicate the system is stable with sufficient margins.

Figure 8.14 Nyquist diagram for , Example 8.4.

Figure 8.15 Bode diagram for , Example 8.4, indicating db and .

Figure 8.16 Bode diagram for open-loop reactor transfer function .

Figure 9.1 Phase volume representing the neutron balance.

Figure 9.2 Lethargy variable and average lethargy increase per collision.

Figure 9.3 Collision density approaching the asymptotic behavior.

Figure 9.4 Collision intervals before and after the collision.

Figure 9.5 Comparison of parameters and as a function of .

Figure 9.6 Illustration of the probability table method for absorption cross section for U over the lethargy interval corresponding to the energy interval eV. Probability density function in the RHS plot represents a distribution of cross section according to its value.

Figure 9.7 Doppler broadening of absorption resonance at energy .

Figure 9.8 Resonance integral function , with , representing Doppler broadening.

Figure 10.1 Weighting factors for perturbations (a) and and (b) .

Figure 10.2 Control rod worth curves: (a) Control rod insertion to position and (b) corresponding rod worth curves.

Figure 10.3 Core and detector geometry for Indian Point Unit 2 plant.

Figure 10.4 Relative assembly power map and excore detector response distribution for the octant of Indian Point Unit 2 core. (a) Relative assembly power map and (b) detector response distribution.

Figure 10.5 Fractional pointwise flux errors in modal expansion calculations for offset fuel assemblies relative to direct 3-D diffusion theory calculations, subject to varying levels of U enrichment perturbations in a fast reactor.

Figure 11.1 Unit-cell representations.

Figure 11.2 Energy dependence of a fuel absorption cross section.

Figure 11.3 Spatial flux distribution as a function of neutron energy.

Figure 11.4 Parameters , , and vs. degree of heterogeneity.

Figure 11.5 Two-region unit cell in slab geometry.

Figure 11.6 Thermal disadvantage factor as a function of fuel half thickness with the ratio kept constant.

Figure 11.7 Flux self-shielding around resonance energy corresponding to for finite dilution factor .

Figure 11.8 Neutron flux spectrum vs. , as a function of .

Figure 11.9 Absorption cross section and normalized flux plots in the thermal range.

Figure 11.10 Comparison of flux spectra for typical SFR and PWR cores.

Figure 11.11 Overall reactor physics calculational procedure.

Figure 12.1 Flow chart for the once-through uranium cycle.

Figure 12.2 Closed fuel cycle with plutonium recycling.

Figure 12.3 Symbiotic two-tier fuel cycle involving both LWR and SFR transmuters.

Figure 12.4 Uranium-plutonium fuel cycle.

Figure 12.5 Thorium-uranium fuel cycle.

Figure 12.6 Extended heavy-nuclide transmutation chain of the CASMO-3 code.

Figure 12.7 Iterative search for an equilibrium cycle.

Figure 12.8 Nuclide vectors for incore and excore fuel cycle processes.

Figure 12.9 Heavy-metal material flow sheet for a closed fuel cycle.

Figure 12.10 One macro-cycle consisting of seven micro-cycles for a cluster of seven driver assemblies in a honeycomb configuration. The number within each assembly represents the number of previous cycles that the assembly has resided in that position. For example, “0” represents a fresh assembly, whereas “6” represents an assembly that has been irradiated for six micro-cycles.

Figure 12.11 Simplified three-region equilibrium cycle modeling.

Figure 12.12 Evolution of I and Xe concentrations during startup and shutdown.

Figure 12.13 Reactivity as a function of fuel burnup.

Figure 12.14 Core reactivity vs. hydrogen-to-U atom ratio.

Figure 12.15 Evolution of power and burnup distributions over a cycle for the Haling power distribution. (a) Power distribution at BOC, (b) burnup Power distribution at EOC, (c) distribution at EOC, and (d) Power distribution at EOC.

Figure 12.16 Material flow for a pressurized water reactor design with once-through LEU fuel. Adapted from NFCE&S Evaluation Group 1.

Figure 12.17 Material flow for a sodium-cooled fast reactor design with once-through HALEU fuel. Adapted from the Natrium Advanced Type 1B Metal core design.

Figure 12.18 Material flow for a high-temperature gas-cooled reactor design with HALEU TRISO fuel. Adapted from NFCE&S Evaluation Group 2.

Figure 12.19 Material flow for a molten salt reactor design with circulating HALEU and thorium fuel. Adapted from NFCE&S Evaluation Group 26.

Figure 12.20 Material flow for a coupled SFR-PWR system, utilizing Pu-U-Zr fuel for the SFR and U-Pu mixed oxide fuel for the PWR. Adapted from NFCE&S Evaluation Group 29.

Figure 12.21 Evolution of UNF ingestion radiotoxicity over time for five representative fuel cycles.

Figure 12.22 Evolution of UNF ingestion radiotoxicity over time for the SKB-3 repository concept.

Figure 12.23 Advanced aqueous reprocessing schemes: homogenous recycling of entire TRU via GANEX and heterogeneous recycling via SANEX that separates Am/Cm from Pu/U and ExAm that separates Am from Cm.

Figure 12.24 UNF radiotoxicity as a function of time over yr.

Figure 12.25 Schematic flowsheet for pyroprocessing of EBR-II metallic fuel.

Figure 12.26 Chemical processes for evolution of species and corrosion of UNF in geological repository.

Figure 12.27 Radionuclides important to Yucca Mount repository mean dose, early (E) and late (L).

Figure 12.28 Major contributors to dose in (a) buffer (top), (b) Opalinus clay (middle), (c) biosphere (bottom) associated with UNF disposal in Switzerland.

Figure 12.29 Overall layout for the Waste Isolation Pilot Plant at Carlsbad, New Mexico.

Figure 12.30 Internal structures of the Onkalo geological repository.

Figure 12.31 UNF Engineered barrier system planned for the Yucca Mountain GR, Nevada.

Figure 13.1 Temperature gradient in a slab.

Figure 13.2 Velocity gradient in fluid between flat plates.

Figure 13.3 Temperature distribution at a solid–fluid interface.

Figure 13.4 Differential fluid volume.

Figure 13.5 Hagen–Poiseuille flow.

Figure 13.6 Shear stress and fluid velocity for the Hagen–Poiseuille flow.

Figure 13.7 Time-dependent one-dimensional velocity profile.

Figure 13.8 Dimensionless velocity profile.

Figure 13.9 Boundary layer thickness as a function of time.

Figure 13.10 Thermal conductivity of at 95% of theoretical density vs. temperature.

Figure 13.11 Heat conduction through three slabs.

Figure 13.12 Radial temperature distribution in a fuel rod cooled by fluid flow.

Figure 13.13 Turbulent velocity profile.

Figure 13.14 Axial temperature profiles in a PWR core.

Figure 13.15 Axial temperature profiles in a BWR core.

Figure 13.16 Pool boiling regimes with heat flux represented as a function of wall superheat .

Figure 13.17 Flow regimes in forced convective flow.

Figure 13.18 Temperature distribution illustrating the onset of local boiling and bulk boiling.

Figure 13.19 Two separate phases in a flow channel.

Figure 13.20 Void fraction vs. flow quality.

Figure 13.21 Void fraction and coolant density variation in a BWR channel.

Figure 13.22 Graphical construct to determine number of fuel rods reaching DNB. (a) DNB probability as a function of the DNB ratio and (b) fuel census curve representing the DNB ratio vs. the corresponding number of rods less than indicated. Radial peaking factor corresponding to is also indicated.

Figure 13.23 Coolant channel illustrating superheated liquid layer with tiny bubbles between bubble layer and heated wall.

Figure 13.24 Iterative determination of MDNBR with accurate estimate of flow quality.

Figure 13.25 Iterative determination of MCPR with accurate estimate of flow quality.

Figure 13.26 Power peaking factor as a function of axial offset for PWR plants.

Figure 13.27 Power capability determination accounting for limiting fuel temperature and heat flux.

Figure 13.28 Staggered mesh structure for control volume .

Figure 13.29 NSSS nodalization diagram for PWR plant.

Figure 14.1 Moderator temperature feedback effects on reactivity.

Figure 14.2 Burnup dependence of reactivity coefficients in LWRs.

Figure 14.4 Reactivity feedback for a ULOF transient in a metallic fuel SFR.

Figure 14.3 Capture-to-fission cross section ratio for Pu.

Figure 15.1 Power capital cost index for North America, 2000–2016.

Figure 15.2 Overall mass balance for the enrichment process.

Figure 15.3 Effective LCOE comparison for various technologies.

Figure 16.1 Reactivity insertion [dollar] arising from control rod ejection from a HZP configuration.

Figure 16.2 Comparison of power transient due to HZP control rod ejection between the PKE solution and quasi-static solution with the PARCS code. Reactivity for the PKE calculation is derived from the PARCS calculation.

Figure 16.3 Overall layout of the KAHTER critical assembly for the pebble-bed reactor.

Figure 16.4 Top view of the instrumentation channel layout for the KAHTER facility. The radii are expressed in mm.

Figure 16.5 Integral rod worth as a function of time after rod insertion without space-time correction.

Figure 16.6 Integral rod worth as a function of time after rod insertion corrected with the modal-local method.

Figure 16.7 Evolution of flux perturbations into perturbations for xenon and iodine concentrations.

Figure 16.8 Determination of system parameter via simulating a control rod bank insertion.

Figure 16.9 Effect of flux perturbations on moderator temperature and reactivity feedback. Plot (a) indicates the flux distribution shifted toward the bottom half of the core, (b) effect of flux perturbation on the moderator temperature (dotted curve) compared with the nominal distribution (solid curve), (c) perturbations in the moderator temperature distribution, and (d) moderator temperature feedback averaged over the top and bottom halves of the core (dotted lines) relative to the core average feedback (solid line).

Figure 16.10 Composite mode representing a combination of the diametral and first azimuthal modes.

Figure 16.11 Xenon-induced power oscillations in the – plane at a three-loop PWR plant.

Figure 16.12 Phase-plane trajectory for the time-optimal shutdown strategy.

Figure 16.13 Time-domain flux program for time-optimal shutdown strategy.

Figure 16.14 Phase-plane trajectory for free-running oscillation and target line for control.

Figure 16.15 Time-optimal trajectories for constrained AO control.

Figure 16.16 Time-domain simulation of optimal control with AO constraints.

Figure 16.17 Model-based controller with control gain and filter gain connected to the system transfer function providing optimal control to the plant , while measurement is fed back to the controller .

Figure 16.18 Augmented plant comprising nominal plant and loop shaping weights and coupled with model based controller (LHS diagram), with indicated in a summary layout (RHS diagram).

Figure 16.19 controller indicating stable control of reactivity and power associated with NCDWOs in a BWR core.

Figure 16.20 LQG regulator for a BWR core indicating tendency of unstable power oscillations.

Figure 17.1 Geometry for a diffusing lump.

Figure 17.2 Two-region unit cell.

Figure 17.3 Geometry for escape probability calculation.

Figure 17.4 Chord length in a lump of volume .

Figure 17.5 Chord length for a semi-infinite slab of thickness .

Figure 17.6 Chord length for a sphere of radius .

Figure 17.7 Escape probability and average flux in a volume immersed in .

Figure 17.8 Lattice arrangement for Dancoff factor evaluation.

Figure 17.9 Infinitely long cylinder for transmission probability calculation.

Figure 17.10 Geometry for collision probability .

Figure 17.11 Unit-cell geometry for collision probability calculation.

Figure 17.12 Relationship between PDF and CDF.

Figure C.1 Gamma function for real .

Figure C.2 Bessel functions of the first kind , and .

Figure C.3 Bessel functions of the second kind .

Figure C.4 Modified Bessel functions and .

Figure D.1 Application of Laplace transform to the ODE solution.

Figure E.1 Virtual displacement separating the actual particle trajectory (solid curve) and non-physical trajectory (dotted curve).

Figure E.2 Total variation at the target trajectory with variable terminal time .

Figure E.3 Feasible trajectories corresponding to .

Figure E.4 Optimal system trajectory and corresponding control for Example E.1.

Figure E.5 Optimal system trajectory in the phase plane and costate variables for Example E.2.

Figure F.1 Flow of information for the Kalman filter.

Preface

The book has been developed to introduce undergraduate and graduate students in nuclear engineering, as well as practicing engineers, to basic concepts of nuclear reactor physics and applications of the concepts to the analysis, design, control, and operation of nuclear reactors. The basic concepts are discussed and the associated mathematical formulations presented with the understanding that the reader has solid background in differential equations and linear algebra. A focus is placed on the use of neutron diffusion theory, with a minimum use of the neutron transport equation, for the development of techniques for lattice physics and global reactor system studies. When the neutron transport equation is used, effort is made to stay with one-dimensional forms of the Boltzmann equation and Legendre polynomials, without invoking the full-blown three-dimensional Boltzmann equation and spherical harmonics. Recent developments in numerical algorithms, including the Krylov subspace method, and the MATLAB software, including the Simulink toolbox, are discussed for efficient studies of steady-state and transient reactor configurations. In addition, nuclear fuel cycle and associated economics analysis are presented together with the application of modern control theory to reactor operation. A self-contained derivation of fluid conservation equations is presented, together with relevant examples, so that the material could be used in a sequence of courses in nuclear reactor physics and engineering to cover thermal-hydraulic analysis of nuclear systems.

The overall structure of the book allows the coverage of fundamental concepts and tools necessary for nuclear reactor physics studies with the first half of the book up to Chapter 10, as it is usually done in a one-semester senior nuclear engineering course at the University of Michigan. Some of the remaining chapters of the book could be covered in a follow-up semester in the undergraduate curriculum or in graduate courses; Chapters 16 and 17 could likely be covered in separate courses dealing with nuclear reactor kinetics and reactor design analysis, respectively, The author sincerely hopes that the book will augment and update several excellent textbooks that have been used in the nuclear science and engineering curriculum in the United States and elsewhere for the first half century of nuclear energy development.