This expanded new edition develops the theory of nuclear reactors from the fundamentals of fission to the operating characteristics of modern reactors. The first half of the book emphasizes reactor criticality analysis and all of the fundamentals that go into modern calculations. Simplified one group diffusion theory models are presented and extended into sophisticated multi-group transport theory models. The second half of the book deals with the two main topics of interest related to operating reactors – reactor kinetics/dynamics, and in-core fuel management. Additional chapters have been added to expand and bring the material up-to-date and include the utilization of more computer codes. Code models and detailed data sets are provided along with example problems making this a useful text for students and researchers wishing to develop an understanding of nuclear power and its implementation in today’s modern energy spectrum. Covers the fundamentals of neutronic analysis for nuclear reactor systems to help understand nuclear reactor theory; Describes the benefits, uses, safety features, and challenges related to implementation of Small Modular Reactors; Provides examples, data sets, and code to assist the reader in obtaining mastery over the subjects.
This book covers the entire spectrum of the science and technology of nuclear reactor systems, from underlying physics, to next generation system applications and beyond. Beginning with neutron physics background and modeling of transport and diffusion, this self-contained learning tool progresses step-by-step to discussions of reactor kinetics, dynamics, and stability that will be invaluable to anyone with a college-level mathematics background wishing to develop an understanding of nuclear power. From fuels and reactions to full systems and plants, the author provides a clear picture of how nuclear energy works, how it can be optimized for safety and efficiency, and why it is important to the future.
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
This book is intended to provide an introduction to the basic principles of nuclear fission reactors for advanced undergraduate or graduate students of physics and engineering. The presentation is also suitable for physicists or engineers who are entering the nuclear power field without previous experience with nuclear reactors. No background knowledge is required beyond that typically acquired in the first two years of an undergraduate program in physics or engineering. Throughout, the emphasis is on explaining why particular reactor systems have evolved in the way they have, without going into great detail about reactor physics or methods of design analysis, which are already covered in a number of excellent specialist texts. The first two chapters serve as an introduction to the basic physics of the atom and the nucleus and to nuclear fission and the nuclear chain reaction. Chapter 3 deals with the fundamentals of nuclear reactor theory, covering neutron slowing down and the spatial dependence of the neutron flux in the reactor, based on the solution of the diffusion equations. The chapter includes a major section on reactor kinetics and control, including'tempera ture and void coefficients and xenon poisoning effects in power reactors. Chapter 4 describes various aspects offuel management and fuel cycles, while Chapter 5 considers materials problems for fuel and other constituents of the reactor. The processes of heat generation and removal are covered in Chapter 6.
Physics of Nuclear Reactors presents a comprehensive analysis of nuclear reactor physics. Editors P. Mohanakrishnan, Om Pal Singh, and Kannan Umasankari and a team of expert contributors combine their knowledge to guide the reader through a toolkit of methods for solving transport equations, understanding the physics of reactor design principles, and developing reactor safety strategies. The inclusion of experimental and operational reactor physics makes this a unique reference for those working and researching nuclear power and the fuel cycle in existing power generation sites and experimental facilities. The book also includes radiation physics, shielding techniques and an analysis of shield design, neutron monitoring and core operations. Those involved in the development and operation of nuclear reactors and the fuel cycle will gain a thorough understanding of all elements of nuclear reactor physics, thus enabling them to apply the analysis and solution methods provided to their own work and research. This book looks to future reactors in development and analyzes their status and challenges before providing possible worked-through solutions. Cover image: Kaiga Atomic Power Station Units 1 - 4, Karnataka, India. In 2018, Unit 1 of the Kaiga Station surpassed the world record of continuous operation, at 962 days. Image courtesy of DAE, India. Includes methods for solving neutron transport problems, nuclear cross-section data and solutions of transport theory Dedicates a chapter to reactor safety that covers mitigation, probabilistic safety assessment and uncertainty analysis Covers experimental and operational physics with details on noise analysis and failed fuel detection
This revised text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. The book begins with fundamental definitions of units and dimensions, thermodynamic variables and the Laws of Thermodynamics progressing to sections on specific applications of the Brayton and Rankine cycles for power generation and projected reactor systems design issues. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play. There have been significant new findings for intercooled systems since the previous edition published and they will be included in this volume. New technology plans for using a Nuclear Air-Brayton as a storage system for a low carbon grid are presented along with updated component sizes and performance criteria for Small Modular Reactors. Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors.
Dynamics and Control of Nuclear Reactors presents the latest knowledge and research in reactor dynamics, control and instrumentation; important factors in ensuring the safe and economic operation of nuclear power plants. This book provides current and future engineers with a single resource containing all relevant information, including detailed treatments on the modeling, simulation, operational features and dynamic characteristics of pressurized light-water reactors, boiling light-water reactors, pressurized heavy-water reactors and molten-salt reactors. It also provides pertinent, but less detailed information on small modular reactors, sodium fast reactors, and gas-cooled reactors. Provides case studies and examples to demonstrate learning through problem solving, including an analysis of accidents at Three Mile Island, Chernobyl and Fukushima Daiichi Includes MATLAB codes to enable the reader to apply the knowledge gained to their own projects and research Features examples and problems that illustrate the principles of dynamic analysis as well as the mathematical tools necessary to understand and apply the analysis Publishers Note: Table 3.1 has been revised and will be included in future printings of the book with the following data: Group Decay Constant, li (sec-1) Delayed Neutron Fraction (bi) 1 0.0124 0.000221 2 0.0305 0.001467 3 0.111 0.001313 4 0.301 0.002647 5 1.14 0.000771 6 3.01 0.000281 Total delayed neutron fraction: 0.0067
An Ultrahigh Temperature Vapor core Reactor (UTVR) system is investigated in this research. The UTVR can be characterized as a thermal, high power density (hundreds of MW/^^), externally-moderated, 235 coupled core, highly-enriched U, circulating-fuel, steady-state, burst power reactor. The investigated reactor system includes two types of fissioning regions: (1) the central Ultrahigh Temperature Vapor Core region (UTVC) which contains a vapor mixture of highly-enriched uranium tetrafluoride (UF^) fuel and a metal fluoride working fluid at an average temperature of 3000 K and an average pressure of 50 atm; and (2) the Boiler COLumn region (BCOL) which contains highly enriched liquid UF^ fuel. The combination of three features differentiates the UTVR from other nuclear reactor concepts. These three features are as follows: 1. the multi-core configuration resulting in a coupled-core system by means of direct neutron transport through the media; 2. the circulating fuel and the associated neutronic and mass flow coupling between the UTVC and boiler cores; and 3. the employment of a two-phase fissioning fuel, i.e., a liquid-vapor combination. Static and dynamic neutronic analysis of this novel system indicates distinct advantages over other existing or conceptual nuclear power systems. These include a unique combination of some very effective inherent negative reactivity feedbacks such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions, and the mass flow coupling feedback between the two types of fissioning cores. Static neutronic analysis is performed using multidimensional discrete ordi nates and Monte Carlo neutron transport codes. Parameters such as the UTVC and boiler column reactivities and reaction rates, core-to-core neutronic coupling coefficients, and neutron lifetimes as a function of vapor core density and boiler core liquid volume are obtained from the static neutronic analysis. The dynamic behavior of the UTVR is examined using a non-linear model, which incorporates circulating-fuel, coupled-core, point reactor kinetics and energetics equations. These equations are solved using a system analysis code. The dynamic analysis indicates that the unique and strong negative reactivity feedbacks of the UTVR are capable of stabilizing the UTVR safely and quickly even when large reactivity insertions are imposed (6p - $ 1.00). The analysis also shows that the system exhibits good dynamic performance even when an inherent negative reactivity feddback is suppressed (e.g., the vapor fuel density power coefficient of reactivity). However, due to the strength of the UTVR's inherent negative reactivity feedbacks, it is found that external reactivity insertions alone are inadequate for bringing about power level changes during normal operations. Additional methods of reactivity control, such as variations in the mass flow rate of the fuel and/or working fluid or variations in the inlet pressure of the fuel/working fluid entering the boiler columns, are needed to achieve the desired power level control.
Classic textbook for an introductory course in nuclear reactor analysis that introduces the nuclear engineering student to the basic scientific principles of nuclear fission chain reactions and lays a foundation for the subsequent application of these principles to the nuclear design and analysis of reactor cores. This text introduces the student to the fundamental principles governing nuclear fission chain reactions in a manner that renders the transition to practical nuclear reactor design methods most natural. The authors stress throughout the very close interplay between the nuclear analysis of a reactor core and those nonnuclear aspects of core analysis, such as thermal-hydraulics or materials studies, which play a major role in determining a reactor design.
