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Thermal Hydraulic Analysis Of Two Phase Closed Thermosyphon Cooling System For New Cold Neutron Source Moderator Of Breazeale Research Reactor At Penn State
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Book Synopsis Thermal Hydraulic Analysis of Two-phase Closed Thermosyphon Cooling System for New Cold Neutron Source Moderator of Breazeale Research Reactor at Penn State by : Melaku Habte
Download or read book Thermal Hydraulic Analysis of Two-phase Closed Thermosyphon Cooling System for New Cold Neutron Source Moderator of Breazeale Research Reactor at Penn State written by Melaku Habte and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Nuclear Thermal Hydraulic and Two-Phase Flow by : Jun Wang
Download or read book Nuclear Thermal Hydraulic and Two-Phase Flow written by Jun Wang and published by Frontiers Media SA. This book was released on 2018-10-11 with total page 120 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear energy is one of the most important clear energy and contributes more than 10% electric power to human society in the past decades of years. The nuclear thermal hydraulic and two-phase flow is one of the basic branches of nuclear technology and provides structure design and safety analysis to the nuclear power reactors. In the new century, the basic theoretical research of thermal hydraulic and two-phase flow, and innovative design for the next generation nuclear power plants (especially for the small modular reactor and molten salt reactor), along with other nuclear branches, constantly support the development of nuclear technology.
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 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 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
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 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 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:
Book Synopsis Proceedings of Nuclear Thermal Hydraulics by : American Nuclear Society. Thermal Hydraulics Division. Meeting
Download or read book Proceedings of Nuclear Thermal Hydraulics written by American Nuclear Society. Thermal Hydraulics Division. Meeting and published by . This book was released on 1990 with total page 828 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis First Proceedings of Nuclear Thermal Hydraulics by : American Nuclear Society. Thermal Hydraulics Division. Meeting
Download or read book First Proceedings of Nuclear Thermal Hydraulics written by American Nuclear Society. Thermal Hydraulics Division. Meeting and published by . This book was released on 1983 with total page 354 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Second Proceedings of Nuclear Thermal Hydraulics Annual Summer Meeting, June 3 - June 7, 1984, New Orleans by : American Nuclear Society. Thermal Hydraulics Division. Meeting
Download or read book Second Proceedings of Nuclear Thermal Hydraulics Annual Summer Meeting, June 3 - June 7, 1984, New Orleans written by American Nuclear Society. Thermal Hydraulics Division. Meeting and published by . This book was released on 1984 with total page 264 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 A Thermal-hydraulic Analysis of the Cooling System for the 500 KW Virginia Polytechnic Institute and State University Reactor by : Ai-Tai Lo
Download or read book A Thermal-hydraulic Analysis of the Cooling System for the 500 KW Virginia Polytechnic Institute and State University Reactor written by Ai-Tai Lo and published by . This book was released on 1982 with total page 140 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I by :
Download or read book Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I written by and published by . This book was released on 2014 with total page 185 pages. Available in PDF, EPUB and Kindle. Book excerpt: This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.
Book Synopsis Thermal Hydraulic Analysis of the University of Massachusetts Lowell Research Reactor by : Anthony L. Stevens
Download or read book Thermal Hydraulic Analysis of the University of Massachusetts Lowell Research Reactor written by Anthony L. Stevens and published by . This book was released on 2002 with total page 230 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis An Assessment of Thermal Hydraulic Analysis Methods for Pressurized Thermal Shock Evaluations by : Eric P. Young
Download or read book An Assessment of Thermal Hydraulic Analysis Methods for Pressurized Thermal Shock Evaluations written by Eric P. Young and published by . This book was released on 2002 with total page 218 pages. Available in PDF, EPUB and Kindle. Book excerpt: Improved methods of determining temperature transients in reactor systems are desired because of recent interest in Pressurized Thermal Shock (PTS) issues. The research presented herein was performed in support of the Nuclear Regulatory Commission's effort to re-evaluate its existing PTS rules. These rules are particularly important to the re-licensing of aging nuclear power plants. The much advanced computational power available to industry may offer a tool that allows the accurate calculation of temperatures inside the reactor vessel while not being inaccessibly expensive. It is proposed that an off-the-shelf Computational Fluid Dynamic (CFD) code, STAR-CD, can be a competitive tool in solving the thermal hydraulic domain of a reactor system. A comparison of the methodology and accuracy of the code types that have been previously used in PTS and one that has not been used extensively, CFD, is provided. A review of the literature shows that computer codes have been validated for solving PTS scenarios. The highly specialized program, REMIX, has been utilized extensively from 1986 to 1991 to interpret accident scenarios in reactor systems. Other programs are also available that can calculate downcomer temperatures including system and CFD type codes. Three codes representing the three different types of programs available are described in detail in the literature review section. Data appropriate for assessing a program's ability to calculate the response of a system to a PTS scenario is available from the current matrix of PTS tests being completed at the APEX-CE facility of the Oregon State University Nuclear Engineering department. The facility is a reduced scale integral test facility originally built for modeling the then-proposed AP-600 plant designed by Westinghouse. For the current test series, the facility was modified to model the Palisades nuclear power plant, a Combustion Engineering Pressurized Water Reactor (PWR). Two of the tests were chosen for their PTS typical conditions to compare with calculations of STAR-CD, REMIX, and RELAP. The computer models in each of the programs were either created, modified from a previous version, or the calculations for the comparisons were contributed. The downcomer temperatures at several locations and cold leg temperature gradients, where available, were extracted from the data and calculations and compared. Comparisons are presented in chapter 5 with graphs, along with some interpretation of the comparisons. It was found that STAR-CD agreed best with the data set in the downcomer and is the only program that calculated the temperature gradient in the cold legs. The agreement of STAR-CD with the cold leg data is also very good. REMIX and RELAP calculations agreement with data for downcomer temperatures are found to be good for all comparisons made, qualitatively more than quantitatively when contrasted with the STAR-CD calculations.
Book Synopsis A Static, Two-phase, Thermal Hydraulic Feedback Model for the Nodal Code QUANDRY by : Hussein Shoukry Khalil
Download or read book A Static, Two-phase, Thermal Hydraulic Feedback Model for the Nodal Code QUANDRY written by Hussein Shoukry Khalil and published by . This book was released on 1980 with total page 40 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis In-vessel Thermal-hydraulic Analysis by :
Download or read book In-vessel Thermal-hydraulic Analysis written by and published by . This book was released on 1983 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper presents some recent results obtained from the COMMIX-1A code for the EBR-II reactor transient No. 10, Phase 2. Both the reactor vessel and the neutron shield assembly and assembly arrangement in the reactor core are shown. The computational grid system used in COMMIX-1A is presented. Reactor flow and power transients are shown. Velocity and temperature distributions at steady state and t (time) = 60 sec are included. Finally, a comparison between the calculated results from COMMIX-1A and experimental measurements are presented for outlet temperatures for driver subassembly of XX08, top-of-core temperature for driver subassembly of XX08, and low-pressure plenum mass flow respectively.
