Nuclear Engineering: Mathematical Modeling and Simulation 1st Edition by Zafar 9780323908313 0323908314

Nuclear Engineering: Mathematical Modeling and Simulation 1st Edition Zafar Ullah Koreshi – Ebook Instant Download/Delivery ISBN(s): 9780323908313, 0323908314

Product details:

• ISBN 10:0323908314
• ISBN 13:9780323908313
• Author: Zafar 

Nuclear Engineering
Mathematical Modeling and Simulation

Table contents:

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

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