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Nuclear Reactor Physics / Edition 2

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Overview

Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developments.

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Editorial Reviews

From the Publisher
From reviews of the first edition:
"...Stacey...developed the material in Part I dealing with basic concepts and theory, in teaching undergraduate and first-year graduate students.... Part II is a more mathematically intense treatment of advanced topics." SciTech Book News

"The first comprehensive book on the subject in 25 years." La Doc STI

"I hope that the publication of this impressive text, whose strength lies in its breadth and it modernity, will accompany a renewed interest in nuclear power expressed through fission reactors." Physics Today

SciTech Book News
...Stacey...developed the material in Part I dealing with basic concepts and theory, in teaching undergraduate and first-year graduate students...Part II is a more mathematically intense treatment of advanced topics.
SciTech Book News
...Stacey...developed the material in Part I dealing with basic concepts and theory, in teaching undergraduate and first-year graduate students...Part II is a more mathematically intense treatment of advanced topics.
Booknews
With roots in the Manhattan Project, nuclear reactor physics is the core discipline of the field of nuclear engineering. Besides nuclear power reactors increasingly supplying global electrical power, nuclear power is also employed in basic physics research, naval propulsion reactors, mobile power sources, and production of radio-isotopes for medical and national security applications. Stacey is with the Georgia Institute of Technology, were he developed the material in Part I dealing with basic concepts and theory, in teaching undergraduate and first-year graduate students. Part II is a more mathematically intense treatment of advanced topics. Includes problems to be worked without an answer key. Appended material includes some useful nuclear data and mathematical formulas; step functions, delta functions, and other "exotic beasts"; and introductions to matrix algebra and Laplace transforms. Annotation c. Book News, Inc., Portland, OR (booknews.com)
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Product Details

  • ISBN-13: 9783527406791
  • Publisher: Wiley
  • Publication date: 7/10/2007
  • Edition description: 2nd, Completely Revised and Enlarged Edition
  • Edition number: 2
  • Pages: 735
  • Sales rank: 1,346,473
  • Product dimensions: 6.99 (w) x 9.65 (h) x 1.64 (d)

Meet the Author

Weston M. Stacey is Professor of Nuclear Engineering at the Georgia Institute of Technology. His career spans more than 40 years of research and teaching in nuclear reactor physics, fusion plasma physics and fusion and fission reactor conceptual design. He led the IAEA INTOR Workshop (1979-88) that led to the present ITER project, for which he was awarded the US Department of Energy Distinguished Associate Award and two Department of Energy Certificates of Appreciation. Professor Stacey is a Fellow of the American Nuclear Society and of the American Physical Society. He is the recipient of several prizes, among them the American Nuclear Society Seaborg Medal for Nuclear Research and the Wigner Reactor Physics Award, and the author of six previous books and numerous research papers.

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Table of Contents

Preface xxiii

Preface to 2nd Edition xxvii

PART 1 BASIC REACTOR PHYSICS

1 Neutron Nuclear Reactions 3

1.1 Neutron-Induced Nuclear Fission 3

1.2 Neutron Capture 13

1.3 Neutron Elastic Scattering 20

1.4 Summary of Cross-Section Data 24

1.5 Evaluated Nuclear Data Files 24

1.6 Elastic Scattering Kinematics 27

2 Neutron Chain Fission Reactors 33

2.1 Neutron Chain Fission Reactions 33

2.2 Criticality 37

2.3 Time Dependence of a Neutron Fission Chain Assembly 38

2.4 Classification of Nuclear Reactors 40

3 Neutron Diffusion Theory 43

3.1 Derivation of One-Speed Diffusion Theory 43

3.2 Solutions of the Neutron Diffusion Equation in Nonmultiplying

3.3 Diffusion Kernels and Distributed Sources in a Homogeneous

3.4 Albedo Boundary Condition 52

3.5 Neutron Diffusion and Migration Lengths 53

3.6 Bare Homogeneous Reactor 57

3.7 Reflected Reactor 62

3.8 Homogenization of a Heterogeneous Fuel–Moderator Assembly 65

3.9 Control Rods 73

3.10 Numerical Solution of Diffusion Equation 77

3.11 Nodal Approximation 83

3.12 Transport Methods 85

4 Neutron Energy Distribution 101

4.1 Analytical Solutions in an Infinite Medium 101

4.2 Multigroup Calculation of Neutron Energy Distribution in an Infinite

4.3 Resonance Absorption 117

4.4 Multigroup Diffusion Theory 127

5 Nuclear Reactor Dynamics 143

5.1 Delayed Fission Neutrons 143

5.2 Point Kinetics Equations 147

5.3 Period–Reactivity Relations 148

5.4 Approximate Solutions of the Point Neutron Kinetics Equations 150

5.5 Delayed Neutron Kernel and Zero-Power Transfer Function 155

5.6 Experimental Determination of Neutron Kinetics Parameters 156

5.7 Reactivity Feedback 161

5.8 Perturbation Theory Evaluation of Reactivity Temperature

5.9 Reactor Stability 171

5.10 Measurement of Reactor Transfer Functions 179

5.11 Reactor Transients with Feedback 183

5.12 Reactor Fast Excursions 186

5.13 Numerical Methods 190

6 Fuel Burnup 197

6.1 Changes in Fuel Composition 197

6.2 Samarium and Xenon 211

6.3 Fertile-to-Fissile Conversion and Breeding 217

6.4 Simple Model of Fuel Depletion 219

6.5 Fuel Reprocessing and Recycling 221

6.6 Radioactive Waste 226

6.7 Burning Surplus Weapons-Grade Uranium and Plutonium 233

6.8 Utilization of Uranium Energy Content 235

6.9 Transmutation of Spent Nuclear Fuel 237

6.10 Closing the Nuclear Fuel Cycle 244

7 Nuclear Power Reactors 249

7.1 Pressurized Water Reactors 249

7.2 Boiling Water Reactors 250

7.3 Pressure Tube Heavy Water–Moderated Reactors 255

7.4 Pressure Tube Graphite-Moderated Reactors 258

7.5 Graphite-Moderated Gas-Cooled Reactors 260

7.6 Liquid-Metal Fast Breeder Reactors 261

7.7 Other Power Reactors 265

7.8 Characteristics of Power Reactors 265

7.9 Advanced Generation-III Reactors 265

7.10 Advanced Generation-IV Reactors 269

7.11 Advanced Sub-critical Reactors 273

7.12 Nuclear Reactor Analysis 275

7.13 Interaction of Reactor Physics and Reactor Thermal Hydraulics 280

8 Reactor Safety 283

8.1 Elements of Reactor Safety 283

8.2 Reactor Safety Analysis 285

8.3 Quantitative Risk Assessment 288

8.4 Reactor Accidents 293

8.5 Passive Safety 299

PART 2 ADVANCED REACTOR PHYSICS

9 Neutron Transport Theory 305

9.1 Neutron Transport Equation 305

9.2 Integral Transport Theory 310

9.3 Collision Probability Methods 319

9.4 Interface Current Methods in Slab Geometry 323

9.5 Multidimensional Interface Current Methods 330

9.6 Spherical Harmonics (PL) Methods in One-Dimensional

9.7 Multidimensional Spherical Harmonics (PL) Transport Theory 350

9.8 Discrete Ordinates Methods in One-Dimensional Slab Geometry 354

9.9 Discrete Ordinates Methods in One-Dimensional Spherical

9.10 Multidimensional Discrete Ordinates Methods 363

9.11 Even-Parity Transport Formulation 369

9.12 Monte Carlo Methods 371

10 Neutron Slowing Down 385

10.1 Elastic Scattering Transfer Function 385

10.2 P1 and B1 Slowing-Down Equations 390

10.3 Diffusion Theory 396

10.4 Continuous Slowing-Down Theory 400

10.5 Multigroup Discrete Ordinates Transport Theory 411

11 Resonance Absorption 415

11.1 Resonance Cross Sections 415

11.2 Widely Spaced Single-Level Resonances in a Heterogeneous Fuel–Moderator Lattice 415

11.3 Calculation of First-Flight Escape Probabilities 424

11.4 Unresolved Resonances 428

11.5 Multiband Treatment of Spatially Dependent Self-Shielding 433

11.6 Resonance Cross-Section Representations 439

12 Neutron Thermalization 453

12.1 Double Differential Scattering Cross Section for Thermal Neutrons 453

12.2 Neutron Scattering from a Monatomic Maxwellian Gas 454

12.3 Thermal Neutron Scattering from Bound Nuclei 457

12.4 Calculation of the Thermal Neutron Spectra in Homogeneous Media 462

12.5 Calculation of Thermal Neutron Energy Spectra in Heterogeneous Lattices 474

12.6 Pulsed Neutron Thermalization 477

13 Perturbation and Variational Methods 483

13.1 Perturbation Theory Reactivity Estimate 483

13.2 Adjoint Operators and Importance Function 486

13.3 Variational/Generalized Perturbation Reactivity Estimate 489

13.4 Variational/Generalized Perturbation Theory Estimates of Reaction Rate Ratios in Critical Reactors 495

13.5 Variational/Generalized Perturbation Theory Estimates of Reaction Rates 497

13.6 Variational Theory 498

13.7 Variational Estimate of Intermediate Resonance Integral 500

13.8 Heterogeneity Reactivity Effects 502

13.9 Variational Derivation of Approximate Equations 503

13.10 Variational Even-Parity Transport Approximations 505

13.11 Boundary Perturbation Theory 508

14 Homogenization 515

14.1 Equivalent Homogenized Cross Sections 516

14.2 ABH Collision Probability Method 517

14.3 Blackness Theory 520

14.4 Fuel Assembly Transport Calculations 522

14.5 Homogenization Theory 529

14.6 Equivalence Homogenization Theory 531

14.7 Multiscale Expansion Homogenization Theory 535

14.8 Flux Detail Reconstruction 538

15 Nodal and Synthesis Methods 541

15.1 General Nodal Formalism 542

15.2 Conventional Nodal Methods 544

15.3 Transverse Integrated Nodal Diffusion Theory Methods 547

15.4 Transverse Integrated Nodal Integral Transport Theory Models 554

15.5 Transverse Integrated Nodal Discrete Ordinates Method 561

15.6 Finite-Element Coarse Mesh Methods 563

15.7 Variational Discrete Ordinates Nodal Method 571

15.8 Variational Principle for Multigroup Diffusion Theory 580

15.9 Single-Channel Spatial Synthesis 583

15.10 Multichannel Spatial Synthesis 589

15.11 Spectral Synthesis 591

16 Space–Time Neutron Kinetics 599

16.1 Flux Tilts and Delayed Neutron Holdback 599

16.2 Spatially Dependent Point Kinetics 602

16.3 Time Integration of the Spatial Neutron Flux Distribution 609

16.4 Stability 625

16.5 Xenon Spatial Oscillations 641

16.6 Stochastic Kinetics 652

APPENDICES

A Physical Constants and Nuclear Data 669

B Some Useful Mathematical Formulas 675

C Step Functions, Delta Functions, and Other Functions 677

C.1 Introduction 677

C.2 Properties of the Dirac δ-Function 678

A. Alternative Representations 678

B. Properties 678

C. Derivatives 679

D Some Properties of Special Functions 681

E Introduction to Matrices and Matrix Algebra 687

E.1 Some Definitions 687

E.2 Matrix Algebra 689

F Introduction to Laplace Transforms 691

F.1 Motivation 691

F.2 “Cookbook” Laplace transforms 694

Index 697

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