Describes the state of knowledge of natural circulation in water cooled nuclear power plants and passive system reliability. The publication presents information on phenomena, models, predictive tools and experiments that currently support design and analysis of natural circulation systems, and highlights areas where additional research is needed.

In recent years it has been recognized that the application of passive safety systems (i.e., those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. In 1991 the IAEA Conference on ''The Safety of Nuclear Power: Strategy for the Future'' noted that for new plants the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate''.

Since the mid-1980s there has been much discussion of the advantages of passive safety systems in advanced nuclear power plants (NPPs). It has been recognized that the application of passive safety systems can contribute to simplification and potentially improve economics of new NPP designs. However, this implies careful design and analysis methods to assure that these systems perform their intended functions. This publication draws on the studies of an IAEA coordinated research project on the topic and reports the findings of the latest research activities. It describes passive safety systems in a wide range of advanced water-cooled nuclear power plant designs, defines the thermal hydraulic phenomena associated with natural circulation phenomena and cross-links these phenomena with the passive safety systems.

Based on an IAEA coordinated research project focused on the use of passive safety systems and natural circulation to help meet the safety and economic goals of advanced nuclear power plants, this publication includes the identification and definition of the thermo-hydraulic phenomena that affect the reliability of passive safety systems, characterization of each phenomenon, integral tests to examine the passive systems and natural circulation, and a methodology for examining passive system reliability.

Single-Phase, Two-Phase and Supercritical Natural Circulation Systems provides readers with a deep understanding of natural circulation systems. This book equips the reader with an understanding on how to detect unstable loops to ensure plant safety and reliability, calculate heat transport capabilities, and design effective natural circulation loops, stability maps and parallel channel systems. Each chapter begins with an introduction to the circulation system before discussing each element in detail and analyzing its effect on the performance of the system. The book also presents thermosyphon heat transport devices in nuclear and other industrial plants, a common information need for students and researchers alike. This book is invaluable for engineers, designers, operators and consultants in nuclear, mechanical, electrical and chemical disciplines. - Presents single-phase, two-phase and supercritical natural circulation systems together in one resource to fill an existing knowledge gap - Guides the reader through relevant processes, such as designing, analyzing and generating stability maps and natural circulation loops, calculating heat transport capabilities, and maintaining natural circulation system operations - Includes global case studies and examples to increase understanding, along with important IAEA standards and procedures

Thermal Hydraulics of Water-Cooled Nuclear Reactors reviews flow and heat transfer phenomena in nuclear systems and examines the critical contribution of this analysis to nuclear technology development. With a strong focus on system thermal hydraulics (SYS TH), the book provides a detailed, yet approachable, presentation of current approaches to reactor thermal hydraulic analysis, also considering the importance of this discipline for the design and operation of safe and efficient water-cooled and moderated reactors. Part One presents the background to nuclear thermal hydraulics, starting with a historical perspective, defining key terms, and considering thermal hydraulics requirements in nuclear technology. Part Two addresses the principles of thermodynamics and relevant target phenomena in nuclear systems. Next, the book focuses on nuclear thermal hydraulics modeling, covering the key areas of heat transfer and pressure drops, then moving on to an introduction to SYS TH and computational fluid dynamics codes. The final part of the book reviews the application of thermal hydraulics in nuclear technology, with chapters on V&V and uncertainty in SYS TH codes, the BEPU approach, and applications to new reactor design, plant lifetime extension, and accident analysis. This book is a valuable resource for academics, graduate students, and professionals studying the thermal hydraulic analysis of nuclear power plants and using SYS TH to demonstrate their safety and acceptability. - Contains a systematic and comprehensive review of current approaches to the thermal-hydraulic analysis of water-cooled and moderated nuclear reactors - Clearly presents the relationship between system level (top-down analysis) and component level phenomenology (bottom-up analysis) - Provides a strong focus on nuclear system thermal hydraulic (SYS TH) codes - Presents detailed coverage of the applications of thermal-hydraulics to demonstrate the safety and acceptability of nuclear power plants

Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 3, Procedures and Applications includes all new chapters which delve deeper into the topic, adding context and practical examples to help readers apply learnings to their own setting. Topics covered include experimental thermal-hydraulics and instrumentation, numerics, scaling and containment in thermal-hydraulics, as well as a title dedicated to good practices in verification and validation. This book will be a valuable reference for graduate and undergraduate students of nuclear or thermal engineering, as well as researchers in nuclear thermal-hydraulics and reactor technology, engineers working in simulation and modeling of nuclear reactors, and more. In addition, nuclear operators, code developers and safety engineers will also benefit from the practical guidance provided. - Presents a comprehensive analysis on the connection between nuclear power and thermal hydraulics - Includes end-of-chapter questions, quizzes and exercises to confirm understanding and provides solutions in an appendix - Covers applicable nuclear reactor safety considerations and design technology throughout

Over the last 30 years, reactor safety technology has evolved not so much from a need to recover from accidents or incidents, but primarily from many groups in the nuclear community asking hypo thetical, searching (what if) ~uestions. This ~uestioning has indeed paid off in establishing preventive measures for many types of events and potential accidents. Conditions, such as reactivity excursions, large break, loss of coolant, core melt, and contain ment integrity loss, to name a few, were all at one time topics of protracted discussions on hypothesized events. Historically, many of these have become multiyear, large-scale research programs aimed at resolving the "what ifs. " For the topic of anticipated and abnormal plant transients, how ever, the searching ~uestions and the research were not so prolific until the mid-1970s. At that time, probabilistic risk methodolo gies began to tell us we should change our emphasis in reactor safety research and development and focus more on small pipe breaks and plant transients. Three Mile Island punctuated that message in 1979. The plant transient topic area is a multidisciplinary subject involving not only the nuclear, fluid flow, and heat transfer technologies, but also the synergistics of these with the reactor control systems, the safety s;,"stems, operator actions, maintenance and even management and the economic considerations of a given plant.

Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain the understanding of the conceptual design elements and specific analysis methods of supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference to engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology.