Monte Carlo Principles and Neutron Transport Problems


Book Description

This two-part treatment introduces the general principles of the Monte Carlo method within a unified mathematical point of view, applying them to problems in neutron transport. It describes several efficiency-enhancing approaches, including the method of superposition and simulation of the adjoint equation based on reciprocity. The first half of the book presents an exposition of the fundamentals of Monte Carlo methods, examining discrete and continuous random walk processes and standard variance reduction techniques. The second half of the text focuses directly on the methods of superposition and reciprocity, illustrating their applications to specific neutron transport problems. Topics include the computation of thermal neutron fluxes and the superposition principle in resonance escape computations.




Neutron Transport


Book Description

This textbook provides a thorough explanation of the physical concepts and presents the general theory of different forms through approximations of the neutron transport processes in nuclear reactors and emphasize the numerical computing methods that lead to the prediction of neutron behavior. Detailed derivations and thorough discussions are the prominent features of this book unlike the brevity and conciseness which are the characteristic of most available textbooks on the subject where students find them difficult to follow. This conclusion has been reached from the experience gained through decades of teaching. The topics covered in this book are suitable for senior undergraduate and graduate students in the fields of nuclear engineering and physics. Other engineering and science students may find the construction and methodology of tackling problems as presented in this book appealing from which they can benefit in solving other problems numerically. The book provides access to a one dimensional, two energy group neutron diffusion program including a user manual, examples, and test problems for student practice. An option of a Matlab user interface is also available.




Mathematical Topics In Neutron Transport Theory: New Aspects


Book Description

This book presents some recent mathematical developments about neutron transport equations. Several different topics are dealt with including regularity of velocity averages, spectral analysis of transport operators, inverse problems, nonlinear problems arising in the stochastic theory of neutron chain fissions, compactness properties of perturbed of c0-semigroups in Banach spaces with applications to transport theory, Miyadera perturbations of c0-semigroups in Banach spaces with applications to singular transport equations, a thorough analysis of the leading eigenelements of transport operators and their approximation, scattering theory. Besides the new problems addressed in this book a unification and extension of the classical spectral analysis of neutron transport equations is given.




An Improved Elastic and Nonelastic Neutron Transport Algorithm for Space Radiation


Book Description

In addition, a high-energy bidirectional neutron source based on a formula by Ranft showed even further improvement of the fluence results over previous results near the front of the water target where diffusion out the front surface is important. Effects of improved interaction cross sections are modest compared with the addition of the high-energy bidirectional source terms.







Stochastic Neutron Transport


Book Description

This monograph highlights the connection between the theory of neutron transport and the theory of non-local branching processes. By detailing this frequently overlooked relationship, the authors provide readers an entry point into several active areas, particularly applications related to general radiation transport. Cutting-edge research published in recent years is collected here for convenient reference. Organized into two parts, the first offers a modern perspective on the relationship between the neutron branching process (NBP) and the neutron transport equation (NTE), as well as some of the core results concerning the growth and spread of mass of the NBP. The second part generalizes some of the theory put forward in the first, offering proofs in a broader context in order to show why NBPs are as malleable as they appear to be. Stochastic Neutron Transport will be a valuable resource for probabilists, and may also be of interest to numerical analysts and engineers in the field of nuclear research.




Mathematical Topics in Neutron Transport Theory


Book Description

This book presents some recent mathematical developments about neutron transport equations. Several different topics are dealt with including regularity of velocity averages, spectral analysis of transport operators, inverse problems, nonlinear problems arising in the stochastic theory of neutron chain fissions, compactness properties of perturbed of 0-semigroups in Banach spaces with applications to transport theory, Miyadera perturbations of c0-semigroups in Banach spaces with applications to singular transport equations, a thorough analysis of the leading eigenelements of transport operators and their approximation, scattering theory. Besides the new problems addressed in this book a unification and extension of the classical spectral analysis of neutron transport equations is given.










Handbook of Nuclear Engineering


Book Description

This is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all levels, this book provides a condensed reference on nuclear engineering since 1958.