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Scaling Approach And Thermal Hydraulic Analysis In The Reactor Cavity Cooling System Of A High Temperature Gas Cooled Reactor And Thermal Jet Mixing In A Sodium Fast Reactor
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Book Synopsis Scaling Approach and Thermal-hydraulic Analysis in the Reactor Cavity Cooling System of a High Temperature Gas-cooled Reactor and Thermal-jet Mixing in a Sodium Fast Reactor by : Olumuyiwa A. Omotowa
Download or read book Scaling Approach and Thermal-hydraulic Analysis in the Reactor Cavity Cooling System of a High Temperature Gas-cooled Reactor and Thermal-jet Mixing in a Sodium Fast Reactor written by Olumuyiwa A. Omotowa and published by . This book was released on 2014 with total page 502 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Water by :
Download or read book Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Water 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 experimental study investigated the thermal hydraulic behavior and boiling mechanisms present in a scaled reactor cavity cooling system (RCCS). The experimental facility reflects a 1/4 scale model of one conceptual design for decay heat removal in advanced GenIV nuclear reactors. Radiant heaters supply up to 25 kW/m2 onto a three parallel riser tube and cooling panel test section assembly, representative of a 5° sector model of the full scale concept. Derived similarity relations have preserved the thermal hydraulic flow patterns and integral system response, ensuring relevant data and similarity among scales. Attention will first be given to the characterization of design features, form and heat losses, nominal behavior, repeatability, and data uncertainty. Then, tests performed in single-phase have evaluated the steady-state behavior. Following, the transition to saturation and subsequent boiling allowed investigations onto four parametric effects at two-phase flow and will be the primary focus area of remaining analysis. Baseline conditions at two-phase flow were defined by 15.19 kW of heated power and 80% coolant inventory, and resulted in semi-periodic system oscillations by the mechanism of hydrostatic head fluctuations. Void generation was the result of adiabatic expansion of the fluid due to a reduction in hydrostatic head pressure, a phenomena similar to flashing. At higher powers of 17.84 and 20.49 kW, this effect was augmented, creating large flow excursions that followed a smooth and sinusoidal shaped path. Stabilization can occur if the steam outflow condition incorporates a nominal restriction, as it will serve to buffer the short time scale excursions of the gas space pressure and dampen oscillations. The influences of an inlet restriction, imposed by an orifice plate, introduced subcooling boiling within the heated core and resulted in chaotic interactions among the parallel risers. The penultimate parametric examined effects of boil-off and inventory loss, where five different stages of natural circulation flow were identified: single-phase heating, transitional nucleate boiling, hydrostatic head fluctuations, stable two-phase flow, and geysering. Finally, the implementation of the model RCCS to a full scale plant was investigated by a multivariate test simulating an hypothetical accident scenario.
Book Synopsis Thermal-Hydraulics of Water Cooled Nuclear Reactors by : Francesco D'Auria
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
Book Synopsis Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air by : Moses A. Muci
Download or read book Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air written by Moses A. Muci and published by . This book was released on 2014 with total page 284 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Energy Research Abstracts written by and published by . This book was released on 1994-02 with total page 484 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Thermal Hydraulic Analysis of a Reduced Scale High Temperature Gas-Cooled Reactor Test Facility and Its Prototype with MELCOR by : Bradley Aaron Beeny
Download or read book Thermal Hydraulic Analysis of a Reduced Scale High Temperature Gas-Cooled Reactor Test Facility and Its Prototype with MELCOR written by Bradley Aaron Beeny and published by . This book was released on 2013 with total page 228 pages. Available in PDF, EPUB and Kindle. Book excerpt: Pursuant to the energy policy act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the Very High Temperature Reactor (VHTR) that will become the Next Generation Nuclear Plant (NGNP). Although plans to build a demonstration plant at Idaho National Laboratories (INL) are currently on hold, a cooperative agreement on HTGR research between the U.S. Nuclear Regulatory Commission (NRC) and several academic investigators remains in place. One component of this agreement relates to validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform HTGR licensing analyses. Because the NRC has used MELCOR for LWR licensing in the past and because MELCOR was recently updated to include gas-cooled reactor physics models, MELCOR is among the system codes of interest in the cooperative agreement. The impetus for this thesis was a code-to-experiment validation study wherein MELCOR computer code predictions were to be benchmarked against experimental data from a reduced-scale HTGR testing apparatus called the High Temperature Test Facility (HTTF). For various reasons, HTTF data is not yet available from facility designers at Oregon State University, and hence the scope of this thesis was narrowed to include only computational studies of the HTTF and its prototype, General Atomics' Modular High Temperature Gas-Cooled Reactor (MHTGR). Using the most complete literature references available for MHTGR design and using preliminary design information on the HTTF, MELCOR input decks for both systems were developed. Normal and off-normal system operating conditions were modeled via implementation of appropriate boundary and inititial conditions. MELCOR Predictions of system response for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) conditions were checked against nominal design parameters, physical intuition, and some computational results available from previous RELAP5-3D analyses at INL. All MELCOR input decks were successfully built and all scenarios were successfully modeled under certain assumptions. Given that the HTTF input deck is preliminary and was based on dated references, the results were altogether imperfect but encouraging since no indications of as yet unknown deficiencies in MELCOR modeling capability were observed. Researchers at TAMU are in a good position to revise the MELCOR models upon receipt of new information and to move forward with MELCOR-to-HTTF benchmarking when and if test data becomes available. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/148182
Book Synopsis Thermal Hydraulic Analysis of University of Wisconsin-Madison's Experimental Air-cooled Reactor Cavity Cooling System Using RELAP5 by : Donghyun Suh
Download or read book Thermal Hydraulic Analysis of University of Wisconsin-Madison's Experimental Air-cooled Reactor Cavity Cooling System Using RELAP5 written by Donghyun Suh and published by . This book was released on 2016 with total page 90 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors by : Francesco D'Auria
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 932 pages. Available in PDF, EPUB and Kindle. Book excerpt: Handbook on Thermal Hydraulics of Water-Cooled Nuclear Reactors, Volume 1, Foundations and Principles 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
Book Synopsis Scaling Analysis for the Pebble Bed of the Very High Temperature Gas-cooled Reactor Thermal Hydraulic Test Facility by : Benjamin L. Nelson
Download or read book Scaling Analysis for the Pebble Bed of the Very High Temperature Gas-cooled Reactor Thermal Hydraulic Test Facility written by Benjamin L. Nelson and published by . This book was released on 2010 with total page 112 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Very High Temperature Reactor (VHTR) has two possible core configurations, a hexagonal prismatic and a pebble bed. It is essential that an experimental facility be built for the validation of computer codes for the safe operation of the VHTR. The scaling of the prismatic core configuration has been analyzed previously for a large break loss of coolant accident. This is a scaling analysis for the pebble bed core configuration. As part of the full scaling analysis, the bottom up scaling of the pebble bed core for pressure drop and radial heat transfer were conducted. Radiation is the dominant form of heat transfer at high temperatures and was scaled using the two methods of treating radiation in a packed bed of spheres. The results of scaling were compared using FLUENT, a computational fluid dynamics code, using the setup, run, and comparison of a 1/80 azimuthally and 1/4 radial full scale prototype and scaled model. The temperature profiles across the core under natural circulation like conditions were determined for both models. The model and prototype temperature profiles had significant variation at the boundary, but only a few degree variation away from the boundary. Additionally, the radiation transport equation and radiation conductivity were compared, and distortions quantified for the FLUENT models.
Book Synopsis Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors by : Francesco D'Auria
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 818 pages. Available in PDF, EPUB and Kindle. Book excerpt: 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
Book Synopsis Thermal-hydraulic Analysis of Gas-cooled Reactor Core Flows by : Amir Keshmiri
Download or read book Thermal-hydraulic Analysis of Gas-cooled Reactor Core Flows written by Amir Keshmiri and published by . This book was released on 2010 with total page 382 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Comparative Assessment of Thermophysical and Thermohydraulic Characteristics of Lead, Lead-Bismuth and Sodium Coolants for Fast Reactors by : IAEA
Download or read book Comparative Assessment of Thermophysical and Thermohydraulic Characteristics of Lead, Lead-Bismuth and Sodium Coolants for Fast Reactors written by IAEA and published by . This book was released on 2002-06-30 with total page 277 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Thermal Hydraulic Analyses for Coupling High Temperature Gas-Cooled Reactor to Hydrogen Plant by : S. Sherman
Download or read book Thermal Hydraulic Analyses for Coupling High Temperature Gas-Cooled Reactor to Hydrogen Plant written by S. Sherman and published by . This book was released on 2006 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed toperform thermal-hydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermalhydraulic and efficiency points of view.
Book Synopsis Scaling Analysis of the Coupled Heat Transfer Process in the High-temperature Gas-cooled Reactor Core by : J. C. Conklin
Download or read book Scaling Analysis of the Coupled Heat Transfer Process in the High-temperature Gas-cooled Reactor Core written by J. C. Conklin and published by . This book was released on 1986 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water by : Michael Joseph Gorman
Download or read book Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water written by Michael Joseph Gorman and published by . This book was released on 2015 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: An existing experimental Reactor Cavity Cooling System using water as the coolant received extensive instrumentation and control upgrades to allow for a thorough investigation into the single-phase flow behavior of the system under a variety of experimental conditions. Base level conditions used a uniform heat flux at a power level appropriately scaled from a benchmark computer simulation of the Gas Turbine Modular Helium Reactor (GT-MHR) using scaling relationships derived by Argonne National Laboratory. Experiments were setup to gauge the effects of flow throttling, non-uniform heat flux profiles, alternate power levels and alternate coolant inventory levels on the flow distribution in the Cooling Panel, and to investigate the relationships between system variables of applied power, the temperature difference across the Cooling Panel ([delta]T) and flowrate. In addition, a single scoping experiment was executed to observe system performance with coolant at the saturation temperature. The system variables proved to have highly linear relationships amongst each other under all experimental conditions. Flow instabilities were observed in the form of counter-phase sinusoidal oscillations of flowrate and [delta]T, the frequency thereof showed a roughly linear relationship with power. Ultrasonic Velocity Profiling (UVP) was used to determine the flow distribution, which increased at the outlet side of the panel with either increased system flowrate or higher heat flux applied to the outlet side, and vice-versa. The effect caused by flowrate changes was the same whether due to a change in power level or throttling, indicating the fluid's momentum is the driving factor. The phenomenon of sudden, high velocity, short duration flow excursions, called geysering, was observed as the system coolant was brought to saturation. This was caused by the trapping of non-condensable gases in the top horizontal section of the flow loop, which in turn brought the flowrate down considerably, increasing residence time and temperature of the coolant in the Cooling Panel. Subsequent rise of saturated coolant to a higher elevation in the hot leg resulted in flashing of the coolant to steam, whose sudden expansion drove the flow excursion. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/155387
Book Synopsis Liquid Metal Cooled Reactors by : International Atomic Energy Agency
Download or read book Liquid Metal Cooled Reactors written by International Atomic Energy Agency and published by . This book was released on 2007 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Presents a survey of worldwide experience gained with fast breeder reactor design, development and operation. Coverage includes state of the art of liquid metal fast reactor development; lead-bismuth cooled (LBC) ship reactor operation experience and LBC fast power reactor development; and treatment and disposal of spent sodium.
Book Synopsis Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System by :
Download or read book Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A & M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the realizable k-epsilon model with two-layer all y+ wall treatment performs better than the other k-epsilon and k-omega turbulence models when compared to the experimental results and the Reynolds stress transport turbulence model results. A scaling analysis was developed to address the distortions introduced by the CFD model in simulating the physical phenomena inside the RCCS system with respect to the full plant configuration. The scaling analysis demonstrated that both the experimental facility and the CFD model achieve a satisfactory resemblance of the main flow characteristics inside the RCCS cavity region, and convection and radiation heat exchange phenomena are properly scaled from the actual plant.
