This book describes the Proceedings of the International Conference on Nuclear Data for Science and Technology held at Jillich in May 1991. The conference was in a series of application oriented nuclear data conferences organized in the past under the auspices of the Nuclear Energy Agency-Nuclear Data Committee (NEANDC) and with the support of the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP). It was the fIrst international conference on nuclear data held in Germany, with the scientific responsibility entrusted to the Institute of Nuclear Chemistry of the Research Centre Jillich. The scientific programme was established by the International Programme Committee in consultation with the International Advisers, and the NEA and IAEA cooperated in the organization. A total of 328 persons from 37 countries and fIve international organizations participated. The scope of these Proceedings extends to a wide range of interdisciplinary topics dealing with measu rement, calculation, evaluation and application of nuclear data, with a major emphasis on numerical data. Both energy and non-energy related applications are considered and due attention is given to some fundamental aspects relevant to the understanding of nuclear data.

Nuclear Fission and Neutron-Induced Fission Cross-Sections is the first volume in a series on Neutron Physics and Nuclear Data in Science and Technology. This volume serves the purpose of providing a thorough description of the many facets of neutron physics in different fields of nuclear applications. This book also attempts to bridge the communication gap between experts involved in the experimental and theoretical studies of nuclear properties and those involved in the technological applications of nuclear data. This publication will be invaluable to those interested in studying nuclear fission and neutron-induced fission cross-sections, as well as other relevant concepts.

NUCLEAR ENGINEERING FUNDAMENTALS is the most modern, up-to-date, and reader friendly nuclear engineering textbook on the market today. It provides a thoroughly modern alternative to classical nuclear engineering textbooks that have not been updated over the last 20 years. Printed in full color, it conveys a sense of awe and wonder to anyone interested in the field of nuclear energy. It discusses nuclear reactor design, nuclear fuel cycles, reactor thermal-hydraulics, reactor operation, reactor safety, radiation detection and protection, and the interaction of radiation with matter. It presents an in-depth introduction to the science of nuclear power, nuclear energy production, the nuclear chain reaction, nuclear cross sections, radioactivity, and radiation transport. All major types of reactors are introduced and discussed, and the role of internet tools in their analysis and design is explored. Reactor safety and reactor containment systems are explored as well. To convey the evolution of nuclear science and engineering, historical figures and their contributions to evolution of the nuclear power industry are explored. Numerous examples are provided throughout the text, and are brought to life through life-like portraits, photographs, and colorful illustrations. The text follows a well-structured pedagogical approach, and provides a wide range of student learning features not available in other textbooks including useful equations, numerous worked examples, and lists of key web resources. As a bonus, a complete Solutions Manual and .PDF slides of all figures are available to qualified instructors who adopt the text. More than any other fundamentals book in a generation, it is student-friendly, and truly impressive in its design and its scope. It can be used for a one semester, a two semester, or a three semester course in the fundamentals of nuclear power. It can also serve as a great reference book for practicing nuclear scientists and engineers. To date, it has achieved the highest overall satisfaction of any mainstream nuclear engineering textbook available on the market today.

Modelling of Nuclear Reactor Multiphysics: From Local Balance Equations to Macroscopic Models in Neutronics and Thermal-Hydraulics is an accessible guide to the advanced methods used to model nuclear reactor systems. The book addresses the frontier discipline of neutronic/thermal-hydraulic modelling of nuclear reactor cores, presenting the main techniques in a generic manner and for practical reactor calculations.The modelling of nuclear reactor systems is one of the most challenging tasks in complex system modelling, due to the many different scales and intertwined physical phenomena involved. The nuclear industry as well as the research institutes and universities heavily rely on the use of complex numerical codes. All the commercial codes are based on using different numerical tools for resolving the various physical fields, and to some extent the different scales, whereas the latest research platforms attempt to adopt a more integrated approach in resolving multiple scales and fields of physics. The book presents the main algorithms used in such codes for neutronic and thermal-hydraulic modelling, providing the details of the underlying methods, together with their assumptions and limitations. Because of the rapidly expanding use of coupled calculations for performing safety analyses, the analysists should be equally knowledgeable in all fields (i.e. neutron transport, fluid dynamics, heat transfer).The first chapter introduces the book's subject matter and explains how to use its digital resources and interactive features. The following chapter derives the governing equations for neutron transport, fluid transport, and heat transfer, so that readers not familiar with any of these fields can comprehend the book without difficulty. The book thereafter examines the peculiarities of nuclear reactor systems and provides an overview of the relevant modelling strategies. Computational methods for neutron transport, first at the cell and assembly levels, then at the core level, and for one-/two-phase flow transport and heat transfer are treated in depth in respective chapters. The coupling between neutron transport solvers and thermal-hydraulic solvers for coarse mesh macroscopic models is given particular attention in a dedicated chapter. The final chapter summarizes the main techniques presented in the book and their interrelation, then explores beyond state-of-the-art modelling techniques relying on more integrated approaches. - Covers neutron transport, fluid dynamics, and heat transfer, and their interdependence, in one reference - Analyses the emerging area of multi-physics and multi-scale reactor modelling - Contains 71 short videos explaining the key concepts and 77 interactive quizzes allowing the readers to test their understanding