Download or read book Development of Mechanistic Modeling Capabilities for Local Neutronically Coupled Flow Induced Instabilities in Advanced Water Cooled Reactors written by and published by . This book was released on 2009 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The major research objectives of this project included the formulation of flow and heat transfer modeling framework for the analysis of flow-induced instabilities in advanced light water nuclear reactors such as boiling water reactors. General multifield model of two-phase flow, including the necessary closure laws. Development of neurton kinetics models compatible with the proposed models of heated channel dynamics. Formulation and encoding of complete coupled neutronics/thermal-hydraulics models for the analysis of spatially-dependent local core instabilities. Computer simulations aimed at testing and validating the new models of reactor dynamics.

Download or read book Modelling of Nuclear Reactor Multi physics written by Christophe Demazière and published by Academic Press. This book was released on 2019-11-19 with total page 370 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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

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 Neutron Response Analysis in Pulsed Coupled core Reactors written by George C. Hopkins and published by . This book was released on 1968 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: In the Los Alamos Coupled Reactor Experiment, some oscillatory deviations from the expected monotonic increase were observed in the rising portion of the passive core neutron density response. In the present study, a generalized multigroup point kinetics model is employed to analyze the experiment. Several approximations with respect to the number of energy groups, the existence of a reflector, and the between-core neutron distribution function were studied. For some approximations, anomalies similar to those experimentally noticed were observed in the passive core response.

Download or read book Numerical Techniques for Coupled Neutronic thermal Hydraulic Nuclear Reactor Calculations written by Curt M. Betts and published by . This book was released on 1994 with total page 498 pages. Available in PDF, EPUB and Kindle. Book excerpt: The solution of coupled neutronic/thermal hydraulic nuclear reactor calculations requires the treatment of the nonlinear feedback induced by the thermal hydraulic dependence of the neutron cross sections. As a result of these nonlinearities, current solution techniques often diverge during the iteration process. These instabilities arise due to the low level of coupling achieved by these methods between the neutronic and thermal hydraulic components. In this work, this solution method is labeled the Decoupled Iteration (DI) method, and this technique is examined in an effort to improve its efficiency and stability. An examination of the DI method also serves to provide insight into the development of more highly coupled iteration methods. After the examination of several possible iteration procedures, two techniques are developed which achieve both a higher degree of coupling and stability. One such procedure is the Outer Iteration Coupling (OIC) method, which combines the outer iteration of the multigroup diffusion calculation with the controlling iteration of the thermal hydraulic calculations. The OIC method appears to be stable for all cases, while maintaining a high level of efficiency. Another iteration procedure developed is the Modified Axial Coupling (MAC) procedure, which couples the neutronic and thermal hydraulic components at the level of the axial position within the coolant channel. While the MAC method does achieve the highest level of coupling and stability, the efficiency of this technique is less than that of the other methods examined. Several characteristics of these coupled calculation methods are examined during the investigation. All methods are shown to be relatively insensitive to thermal hydraulic operating conditions, while the dependence upon convergence criteria is quite significant. It is demonstrated that the DI method does not converge for arbitrarily small convergence criteria, which is a result of a non-asymptotic solution approximation by the DI method. This asymptotic quality is achieved in the coupled methods. Thus, not only do the OIC and MAC techniques converge for small values of the relevant convergence criteria, but the computational expense of these methods is a predictable function of these criteria. The degree of stability of the iterative techniques is enhanced by a higher level of coupling, but the efficiency of these methods tends to decrease as a higher degree of coupling is achieved. This is apparent in the diminished efficiency of the MAC procedure. Seeking an optimum balance of efficiency and stability, the OIC technique is demonstrated to be the optimum method for coupled neutronic/thermal hydraulic reactor calculations.

Download or read book Single and Two phase Flow Modeling for Coupled Neutronics thermal hydraulics Transient Analysis of Advanced Sodium cooled Fast Reactors written by Aurélia Chenu and published by . This book was released on 2011 with total page 237 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Development of Modeling Capabilities for the Analysis of Supercritical Water Cooled Reactor Thermal Hydraulics and Dynamics written by and published by . This book was released on 2009 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Develop an experimental and theoretical data base for heat transfer in tubes and channels cooled by water and CO2 at supercritical pressures.

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 Low order Multiphysics Coupling Techniques for Nuclear Reactor Applications written by Erik Daniel Walker and published by . This book was released on 2017 with total page 118 pages. Available in PDF, EPUB and Kindle. Book excerpt: The accurate modeling and simulation of nuclear reactor designs depends greatly on the ability to couple differing sets of physics together. Current coupling techniques most often use a fixed-point, or Picard, iteration scheme in which each set of physics is solved separately, and the resulting solutions are passed between each solver. In the work presented here, two different coupling techniques are investigated: a Jacobian-Free Newton-Krylov (JFNK) approach and a new methodology called Coarse Mesh Finite Difference Coupling (CMFD-Coupling). What both of these techniques have in common is that they are applied to the low-order CMFD system of equations. This allows for the multiphysics feedback effects to be captured on the low-order system without having to perform a neutron transport solve.The JFNK and CMFD-Coupling approaches were implemented in the MPACT (Michigan Parallel Analysis based on Characteristic Tracing) neutron transport code, which is being developed for the Consortium for Advanced Simulation of Light Water Reactors (CASL). These methods were tested on a wide range of practical reactor physics problems, from a 2D pin cell to a massively parallel 3D full core problem. Initially, JFNK was implemented only as an eigenvalue solver without any feedback enabled. However this led to greatly increased runtimes without any obvious benefit. When multiphysics problems were investigated with both JFNK and CMFD-Coupling, it was concluded that CMFD-Coupling outperformed JFNK in terms of both accuracy and runtime for every problem. When applied to large full core problems with multiple sources of strong feedback enabled, CMFD-Coupling reduced the overall number of transport sweeps required for convergence.

Download or read book A Theoretical Model for Coupled Neutronic thermohydraulic Out of phase Oscillations in Boiling Water Reactors written by D. D. B. van Bragt and published by . This book was released on 1995 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Multidimensional Reactor Kinetics Modeling written by and published by . This book was released on 1996 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: There is general agreement that for many light water reactor transient calculations, it is-necessary to use a multidimensional neutron kinetics model coupled to a thermal-hydraulics model for satisfactory results. These calculations are needed for a variety of applications for licensing safety analysis, probabilistic risk assessment (PRA), operational support, and training. The latter three applications have always required best-estimate models, but in the past applications for licensing could be satisfied with relatively simple models. By using more sophisticated best-estimate models, the consequences of these calculations are better understood, and the potential for gaining relief from restrictive operating limits increases. Hence, for all of the aforementioned applications, it is important to have the ability to do best-estimate calculations with multidimensional neutron kinetics models. coupled to sophisticated thermal-hydraulic models. Specifically, this paper reviews the status of multidimensional neutron kinetics modeling which would be used in conjunction with thermal-hydraulic models to do core dynamics calculations, either coupled to a complete NSSS representation or in isolation. In addition, the paper makes recommendations as to what should be the state-of-the-art for the next ten years. The review is an update to a previous review of the status as of ten years ago. The general requirements for a core dynamics code and the modeling available for such a code, discussed in that review, are still applicable. The emphasis in the current review is on the neutron kinetics assuming that the necessary thermal-hydraulic capability exists. In addition to discussing the basic neutron kinetics, discussion is given of related modeling (other than thermal- hydraulics). The capabilities and limitations of current computer codes are presented to understand the state-of-the-art and to help clarify the future direction of model development in this area.

Download or read book INIS Atomindex written by and published by . This book was released on 1996 with total page 794 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Status of Research and Technology Development for Supercritical Water Cooled Reactors written by International Atomic Energy Agency and published by International Atomic Energy Agency. This book was released on 2019-05-22 with total page 74 pages. Available in PDF, EPUB and Kindle. Book excerpt: There is considerable interest in both developing and developed countries in the design of innovative water cooled reactors (WCRs) and, owing to the higher thermal efficiency and significant system simplifications, supercritical water cooled reactors (SWCRs). Compared to conventional WCRs the SCWR concept requires extensive, comprehensive research and development (R&D). Fundamental research in understanding important phenomena has been completed successfully in providing information required for the next step of development. Currently, a few concepts have been assessed as being technical feasible, and several other concepts are under development. These concepts are described in this publication, together with detailed analysis of remaining gaps requiring future R&D.

Download or read book Transactions of the American Nuclear Society written by American Nuclear Society and published by . This book was released on 1984 with total page 1088 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Structural Materials for Generation IV Nuclear Reactors written by Pascal Yvon and published by Woodhead Publishing. This book was released on 2016-08-27 with total page 686 pages. Available in PDF, EPUB and Kindle. Book excerpt: Operating at a high level of fuel efficiency, safety, proliferation-resistance, sustainability and cost, generation IV nuclear reactors promise enhanced features to an energy resource which is already seen as an outstanding source of reliable base load power. The performance and reliability of materials when subjected to the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors are essential areas of study, as key considerations for the successful development of generation IV reactors are suitable structural materials for both in-core and out-of-core applications. Structural Materials for Generation IV Nuclear Reactors explores the current state-of-the art in these areas. Part One reviews the materials, requirements and challenges in generation IV systems. Part Two presents the core materials with chapters on irradiation resistant austenitic steels, ODS/FM steels and refractory metals amongst others. Part Three looks at out-of-core materials. Structural Materials for Generation IV Nuclear Reactors is an essential reference text for professional scientists, engineers and postgraduate researchers involved in the development of generation IV nuclear reactors. - Introduces the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors and implications for structural materials - Contains chapters on the key core and out-of-core materials, from steels to advanced micro-laminates - Written by an expert in that particular area

Download or read book Accident Analysis for Nuclear Power Plants written by International Atomic Energy Agency and published by . This book was released on 2002 with total page 144 pages. Available in PDF, EPUB and Kindle. Book excerpt: Accident analysis is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). The purpose of the report is to provide the necessary practical guidance for performing adequate accident analysis in the light of current good practice worldwide.

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