Download or read book Passive Safety Systems and Natural Circulation in Water Cooled Nuclear Power Plants written by and published by . This book was released on 2009 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.
Download or read book Natural Circulation in Water Cooled Nuclear Power Plants written by International Atomic Energy Agency and published by IAEA. This book was released on 2005 with total page 656 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.
Download or read book Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors written by International Atomic Energy Agency and published by . This book was released on 2012 with total page 423 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.
Download or read book Passive Safety Systems in Advanced Water Cooled Reactors Awcrs Case Studies written by International Atomic Energy Agency and published by International Atomic Energy Agency. This book was released on 2013-10-31 with total page 101 pages. Available in PDF, EPUB and Kindle. Book excerpt: "This publication is the final report of the INPRO collaborative project on advanced water cooled reactor (AWCR) case studies in support of passive safety systems. It compiles case study results on natural circulation and thermal stratification phenomena in selected passive safety systems performed by the participating countries of Argentina, India and Republic of Korea. The publication presents case study results as well as major findings and conclusions drawn from these case studies. This publication will be a useful resource for researchers in the area of core physics and reactor thermohydraulics and for technical experts and engineers engaged in experimental investigation and theoretical simulation for developing passive safety systems."--Publisher's description.
Download or read book Natural Circulation in Water Cooled Nuclear Power Plants Phenomena Models and Methodology for System Reliability Assessments written by and published by . This book was released on 2005 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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''.
Download or read book Passive Safety Systems in Advanced Water Cooled Reactors AWCRs written by and published by . This book was released on 2013 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: "This publication is the final report of the INPRO collaborative project on advanced water cooled reactor (AWCR) case studies in support of passive safety systems. It compiles case study results on natural circulation and thermal stratification phenomena in selected passive safety systems performed by the participating countries of Argentina, India and Republic of Korea. The publication presents case study results as well as major findings and conclusions drawn from these case studies. This publication will be a useful resource for researchers in the area of core physics and reactor thermohydraulics and for technical experts and engineers engaged in experimental investigation and theoretical simulation for developing passive safety systems."--Publisher's description.
Download or read book Thermal Hydraulics of Water Cooled Nuclear Reactors written by Francesco D'Auria and published by Woodhead Publishing. This book was released on 2017-05-18 with total page 1200 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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
Download or read book Handbook on Thermal Hydraulics in Water Cooled Nuclear Reactors written by Francesco D'Auria and published by Elsevier. This book was released on 2024-07-29 with total page 1012 pages. Available in PDF, EPUB and Kindle. Book excerpt: Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 2, Modelling 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
Download or read book Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors written by Cristhian Galvez and published by . This book was released on 2011 with total page 508 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data. This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study. A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the transient event and maintains the core and reactor components at acceptable temperatures within allowable safety margins. It is also demonstrated that the transition from steady full-power, forced-cooling mode to steady decay-heat, natural-circulation mode is stable, predictable and well characterized.
Download or read book Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features written by Y. Hayashi and published by . This book was released on 2003 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.
Download or read book Experimental Validation of Passive Safety System Models written by Nicolas Zweibaum and published by . This book was released on 2015 with total page 231 pages. Available in PDF, EPUB and Kindle. Book excerpt: The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). The capability to validate integral transient response models is a key issue for licensing new reactor designs. This dissertation presents the scaling strategy, design and fabrication aspects, and startup testing results from the Compact Integral Effects Test (CIET) facility at UCB, which reproduces the thermal hydraulic response of an FHR under forced and natural circulation operation. CIET provides validation data to confirm the performance of the direct reactor auxiliary cooling system (DRACS) in an FHR, used for natural-circulation-driven decay heat removal, under a set of reference licensing basis events, as predicted by best-estimate codes such as RELAP5-3D. CIET uses a simulant fluid, Dowtherm A oil, which at relatively low temperatures (50-120°C) matches the Prandtl, Reynolds, Froude and Grashof numbers of the major liquid salts simultaneously, at approximately 50% geometric scale and heater power under 2% of prototypical conditions. The studies reported here include isothermal pressure drop tests performed during startup testing of CIET, with extensive pressure data collection to determine friction losses in the system, as well as subsequent heated tests, from parasitic heat loss tests to more complex feedback control tests and natural circulation experiments. For initial code validation, coupled steady-state single-phase natural circulation loops and simple forced cooling transients were conducted in CIET. For various heat input levels and temperature boundary conditions, fluid mass flow rates and temperatures were compared between RELAP5- 3D results, analytical solutions when available, and experimental data. This study shows that RELAP5-3D provides excellent predictions of steady-state natural circulation and simple transient forced cooling in CIET. The code predicts natural circulation mass flow rates within 8%, and steady-state and transient fluid temperatures, under both natural and forced circulation, within 2°C of experimental data, suggesting that RELAP5-3D is a good EM to use to design and license FHRs. A key element in design and licensing of new reactor technology lies in the analysis of the plant response to a variety of potential transients. When applicable, this involves understanding of passive safety system behavior. This dissertation develops a framework to assess reliability and propose design optimization and risk mitigation strategies associated with passive decay heat removal systems, applied to the Mk1 PB-FHR DRACS. This investigation builds upon previous detailed design work for Mk1 components and the use of RELAP5-3D models validated for FHR natural circulation phenomenology. For risk assessment, reliability of the point design of the passive safety system for the Mk1 PB-FHR, which depends on the ability of various structures to fulfill their safety functions, is studied. Whereas traditional probabilistic risk assessment (PRA) methods are based on event and fault trees for components of the system that perform in a binary way - operating or not operating -, this study is mostly based on probability distributions of heat load compared to the capacity of the system to remove heat, as recommended by the reliability methods for passive safety functions (RMPS) that are used here. To reduce computational time, the use of response surfaces to describe the system in a simplified manner, in the context of RMPS, is also demonstrated. The design optimization and risk mitigation part proposes a framework to study the elements of the design of the reactor, and more specifically its passive safety cooling system, which can contribute to enhanced reliability of heat removal under accident conditions. Risk mitigation measures based on design, startup testing, in-service inspection and online monitoring are proposed to narrow probability distributions of key parameters of the system and increase reliability and safety. Another major aspect in the development of novel energy systems is the assessment of their impacts on the environment compared to current technologies. While most existing life cycle assessment (LCA) studies have been applied to conventional nuclear power plants, this dissertation proposes a framework to extend such studies to advanced reactor designs, using the example of the Mk1 PB-FHR. The Mk1 uses a nuclear air-Brayton combined cycle designed to produce 100 MWe of base-load electricity when operated with only nuclear heat, and 242 MWe using natural gas co-firing for peaking power. The Mk1 design provides a basis for quantities and costs of major classes of materials involved in building the reactor and fabricating fuel, and operation parameters. Existing data and economic input-output LCA models are used to calculate greenhouse gas emissions per kWh of electricity produced over the life cycle of the reactor. Baseline life cycle emissions from the Mk1 PB-FHR in base-load configuration are 26% lower than average Generation II light water reactors in the U.S., 98% lower than average U.S. coal plants and 96% lower than average U.S. natural gas combined cycle plants using the same turbine technology. In peaking configuration, due to its nuclear component and higher thermal efficiency, the Mk1 plant only produces 32% of the emissions of average U.S. gas turbine simple cycle peaking plants. One key contribution to life cycle emissions results from the amount and type of concrete used for reactor construction. This is an incentive to develop innovative construction methods using optimized steel-concrete composite wall modules and new concrete mixes to reduce life cycle emissions from the Mk1 and other advanced reactor designs.
Download or read book Nuclear Power written by National Research Council and published by National Academies Press. This book was released on 1992-02-01 with total page 234 pages. Available in PDF, EPUB and Kindle. Book excerpt: The construction of nuclear power plants in the United States is stopping, as regulators, reactor manufacturers, and operators sort out a host of technical and institutional problems. This volume summarizes the status of nuclear power, analyzes the obstacles to resumption of construction of nuclear plants, and describes and evaluates the technological alternatives for safer, more economical reactors. Topics covered include: Institutional issues-including regulatory practices at the federal and state levels, the growing trends toward greater competition in the generation of electricity, and nuclear and nonnuclear generation options. Critical evaluation of advanced reactors-covering attributes such as cost, construction time, safety, development status, and fuel cycles. Finally, three alternative federal research and development programs are presented.
Download or read book Innovative Small and Medium Sized Reactors written by International Atomic Energy Agency and published by IAEA. This book was released on 2005 with total page 232 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication presents a variety of innovative water cooled, gas cooled, liquid metal cooled and non-conventional small and medium sized reactor (SMR) designs developed worldwide, and examines technology and infrastructure development needs that may be common to several concepts or lines of such reactors. It also gives an updated definition of small reactors without on-site refuelling and a preliminary review of the passive safety design options for SMRs.
Download or read book Reactor Coolant and Associated Systems in Nuclear Power Plants written by International Atomic Energy Agency and published by . This book was released on 1986 with total page 96 pages. Available in PDF, EPUB and Kindle. Book excerpt: Covers the safety design considerations for various reactor coolant and associated systems for operational states and accident conditions including the selection, sizing and reliability aspects. This includes safety systems such as emergency core cooling, residual heat removal or emergency feedwater systems.
Download or read book Super Light Water Reactors and Super Fast Reactors written by Yoshiaki Oka and published by Springer Science & Business Media. This book was released on 2010-07-01 with total page 664 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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.
Download or read book Natural Circulation Data and Methods for Advanced Water Cooled Nuclear Power Plant Designs written by International Atomic Energy Agency and published by . This book was released on 2002 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors written by International Atomic Energy Agency and published by IAEA Tecdoc. This book was released on 2014-09 with total page 223 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication summarizes the results of an IAEA coordinated research project on the development of advanced methodologies for the assessment of passive safety system performance in advanced reactors. This includes discussions on various methodologies to assess the performance of passive engineered safety features in innovative small reactors, including the Indian AHWR 300 LEU and the Argentinian CAREM25. The publication focuses on the different reliability assessment approaches, methodologies, analysis and evaluation of the results and technical challenges. It provides the insights resulting from the analysis on the technical issues associated with assessing the reliability of passive systems in the context of nuclear safety and probabilistic safety analysis. A viable path towards the implementation of the research efforts in the related areas is also delineated.
