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Nuclear Engineering
Mathematical Modeling and Simulation
- 1st Edition - March 22, 2022
- Author: Zafar Ullah Koreshi
- Language: English
- Paperback ISBN:9 7 8 - 0 - 3 2 3 - 9 0 6 1 8 - 0
- eBook ISBN:9 7 8 - 0 - 3 2 3 - 9 0 8 3 1 - 3
Nuclear Engineering Mathematical Modeling and Simulation presents the mathematical modeling of neutron diffusion and transport. Aimed at students and early career engineers… Read more
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Request a sales quoteNuclear Engineering Mathematical Modeling and Simulation presents the mathematical modeling of neutron diffusion and transport. Aimed at students and early career engineers, this highly practical and visual resource guides the reader through computer simulations using the Monte Carlo Method which can be applied to a variety of applications, including power generation, criticality assemblies, nuclear detection systems, and nuclear medicine to name a few. The book covers optimization in both the traditional deterministic framework of variational methods and the stochastic framework of Monte Carlo methods.
Specific sections cover the fundamentals of nuclear physics, computer codes used for neutron and photon radiation transport simulations, applications of analyses and simulations, optimization techniques for both fixed-source and multiplying systems, and various simulations in the medical area where radioisotopes are used in cancer treatment.
- Provides a highly visual and practical reference that includes mathematical modeling, formulations, models and methods throughout
- Includes all current major computer codes, such as ANISN, MCNP and MATLAB for user coding and analysis
- Guides the reader through simulations for the design optimization of both present-day and future nuclear systems
Graduates on energy/mechanical/nuclear courses’; researchers and early career researchers involved with neutron physics with knowledge of basic mathematics. Textbook on Nuclear/Mechanical/Energy or Physics courses
- Cover image
- Title page
- Table of Contents
- Copyright
- Dedication
- About the author
- Foreword
- Chapter 1. The atom and nuclear radiation
- Abstract
- 1.1 The atom
- 1.2 Radioactive decay
- 1.3 Interaction of radiation with matter
- 1.4 Sources and effects of radiation
- 1.5 Atomic densities of elements and mixtures
- 1.6 Mathematical modeling and simulation
- Capabilities developed
- Nomenclature
- Problems
- References
- Chapter 2. Interactions of neutrons with matter
- Abstract
- 2.1 Kinetic theory
- 2.2 Types of neutron interactions
- 2.3 The microscopic cross-section
- 2.4 The macroscopic cross-section
- 2.5 Flux measurement
- 2.6 Reaction rates
- 2.7 Neutron slowing down, diffusion and thermalization
- 2.8 Resonance cross-section
- 2.9 Nuclear fission
- 2.10 Criticality
- Problems
- Nomenclature
- References
- Chapter 3. Nuclear reactors and systems
- Abstract
- 3.1 Status of nuclear power
- 3.2 Nuclear reactor systems
- 3.3 Marine propulsion reactors
- 3.4 Plutonium production reactors
- 3.5 Small modular reactors
- 3.6 Nuclear fusion
- 3.7 Space propulsion
- 3.8 Nuclear power systems in space
- 3.9 Conclusions
- Problems
- Nomenclature
- References
- ANNEX: the physics of nuclear fusion
- Chapter 4. Mathematical foundations
- Abstract
- 4.1 Ordinary differential equations
- 4.2 Partial differential equations
- 4.3 Integral equations
- 4.4 Integro-differential equations
- 4.5 Numerical methods
- 4.6 Approximate methods
- 4.7 The adjoint function
- 4.8 Random processes, probability, and statistics
- 4.9 Evaluation of integrals
- Problems
- Nomenclature
- References
- Chapter 5. The neutron diffusion equation
- Abstract
- 5.1 The conservation equation
- 5.2 The one-group diffusion equation
- 5.3 The two-group diffusion equation
- 5.4 The multigroup diffusion equation
- 5.5 Effect of fuel concentration on critical mass
- 5.6 The two-group adjoint diffusion equations
- 5.7 Core neutronics with diffusion equations
- Problems
- Nomenclature
- References
- Chapter 6. The neutron transport equation
- Abstract
- 6.1 Structure of the neutron transport equation
- 6.2 Exact solutions of the transport equation
- 6.3 Numerical methods for solving the transport equation
- 6.4 Transport theory for reactor calculations
- Problems
- Nomenclature
- References
- Chapter 7. The Monte Carlo method
- Abstract
- 7.1 Stochastic simulation
- 7.2 Simulation of a random walk
- 7.3 Modeling the geometry
- 7.4 Demonstration
- 7.5 Variance reduction methods
- 7.6 Estimating perturbations with Monte Carlo simulation
- 7.7 Conclusions
- Problems
- Nomenclature
- References
- Chapter 8. Computer codes
- Abstract
- 8.1 Neutron and radiation transport codes
- 8.2 Time-dependent reactor kinetics codes
- 8.3 Thermal hydraulics codes
- 8.4 Radiological protection codes
- 8.5 Performance and safety analyses
- 8.6 Nuclear data
- 8.7 Conclusion
- Problems
- Nomenclature
- References
- Chapter 9. Optimization and variational methods
- Abstract
- 9.1 Introduction
- 9.2 Deterministic optimization
- 9.3 Controller design and optimization
- 9.4 Dynamic programming
- 9.5 Stochastic optimization
- 9.6 Applications of optimization in reactors
- Problems
- Nomenclature
- References
- Chapter 10. Monte Carlo simulation in nuclear systems
- Abstract
- 10.1 Introduction
- 10.2 Bare critical assemblies
- 10.3 Criticality safety
- 10.4 Radiation moderation and shielding
- 10.5 Nuclear fission applications
- 10.6 Nuclear fusion applications
- Problems
- Nomenclature
- References
- Annex A MCNP listing for Godiva (Section 10.2.1)
- Annex B MCNP input listing (Jezebel, Section 10.2.2)
- Annex C MCNP input listing (BK10Shld, Section 10.5.1)
- Annex D MCNP input listing (BK10AP10, Section 10.5.1)
- Chapter 11. Comparisons: Monte Carlo, diffusion, and transport
- Abstract
- 11.1 Introduction
- 11.2 Criticality in a bare sphere
- 11.3 The classic albedo calculation
- 11.4 Flux in a slab
- 11.5 Flux in a finite sphere with a point isotropic source
- Problems
- Nomenclature
- References
- Annex A MATLAB Program AlbedoSlabDiffTh.m (Section 11.3)
- Annex B MCNP Input File BK11Albd (Section 11.2)
- Annex C MATLAB Program CH11ExactSolSlabJan03.m (Section 11.4.4)
- Chapter 12. Exercises in Monte Carlo simulation
- Abstract
- 12.1 Sampling from a distribution function
- 12.2 Estimating the neutron flux in a non-multiplying sphere
- 12.3 Reflected assemblies
- 12.4 Reactor core modeling
- 12.5 Radiation safety and shielding
- 12.6 Perturbation calculations
- 12.7 MCNP geometry plotting in core neutronics
- Conclusions
- Nomenclature
- References
- Annex A MATLAB Program CH12_NormalSampling.m
- Annex B MATLAB Program CH12_Watt Sampling.m
- Chapter 13. Optimization in nuclear systems
- Abstract
- 13.1 Introduction
- 13.2 Reactor core design optimization
- 13.3 Fusion neutronics design optimization
- 13.4 Radiation shielding design optimization
- 13.5 Fuel loading pattern optimization
- 13.6 Radiation detection or optimization
- 13.7 Controller design optimization
- Problems
- Nomenclature
- References
- Chapter 14. Monte Carlo simulation in medical physics
- Abstract
- 14.1 Introduction
- 14.2 Brachytherapy
- Nomenclature
- References
- Index
- No. of pages: 548
- Language: English
- Edition: 1
- Published: March 22, 2022
- Imprint: Academic Press
- Paperback ISBN: 9780323906180
- eBook ISBN: 9780323908313
ZK
Zafar Ullah Koreshi
Zafar ullah Koreshi [B.Sc. (Hons) Nuclear Engineering, Queen Mary College, University of London (UK); M.S, Nuclear Engineering, University of Wisconsin, Madison (USA), Ph.D Nuclear Engineering, University of Cambridge] is Professor at Air University, having contributed as Dean Faculty of Engineering and currently is Dean Graduate Studies at Air University. His experience has been in the Pakistan Atomic Energy Commission, Dr. A Q Khan Research Laboratories, National University of Sciences and Technology, and at Air University, Islamabad. He has published in Annals of Nuclear Energy, Progress in Nuclear Energy, Nuclear Technology and Radiation Protection, ASME Journal of Nuclear Engineering and Radiation Sciences and at several other leading international journals. Dr. Koreshi has been Track Chair at the International Conference on Nuclear Engineering held in the USA, China and Japan and has presented his research at the American Nuclear Society Annual Meetings. He is Member American Society of Mechanical Engineers (ASME), Professional Engineer Pakistan Engineering Council and Life Member Pakistan Nuclear Society. He has also received commendations for being Reviewer of prestigious journals. Prof. Zafar Koreshi is an Associate Editor of the ASME Journal of Nuclear Engineering and Radiation Science.
Affiliations and expertise
Professor and Dean Graduate Studies, Air University, Islamabad, PakistanRead Nuclear Engineering on ScienceDirect