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Thermal Hydraulic Analysis Of University Of Wisconsin Madisons Experimental Air Cooled Reactor Cavity Cooling System Using Relap5
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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 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 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 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 Onset of Boiling and Two-phase Flow Behavior Analysis of an UW Water-cooled Reactor Cavity Cooling System with RELAP5 by : Sowun Kim
Download or read book Onset of Boiling and Two-phase Flow Behavior Analysis of an UW Water-cooled Reactor Cavity Cooling System with RELAP5 written by Sowun Kim and published by . This book was released on 2014 with total page 238 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Experimental and Computational Study of a Scaled Reactor Cavity Cooling System by : Rodolfo Vaghetto
Download or read book Experimental and Computational Study of a Scaled Reactor Cavity Cooling System written by Rodolfo Vaghetto and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Very High Temperature Gas-Cooled Reactor (VHTR) is one of the next generation nuclear reactors designed to achieve high temperatures to support industrial applications and power generation. The Reactor Cavity Cooling System (RCCS) is a passive safety system that will be incorporated in the VTHR, designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The overall behavior of the facility met the expectations. The steady-state condition was achieved and the facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation. The experimental data produced during the steady-state run were successfully compared with the simulation results obtained using RELAP5-3D, confirming the capabilities of the system code of simulating the thermal-hydraulic phenomena occurring in the reactor cavity. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/151742
Book Synopsis Heat Transfer Simulation of Reactor Cavity Cooling System Experimental Facility Using RELAP5-3D and Generation of View Factors Using MCNP by : Huali Wu
Download or read book Heat Transfer Simulation of Reactor Cavity Cooling System Experimental Facility Using RELAP5-3D and Generation of View Factors Using MCNP written by Huali Wu and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: As one of the most attractive reactor types, the High Temperature Gas-cooled Reactor (HTGR) is designed to be passively safe with the incorporation of Reactor Cavity Cooling System (RCCS). In this paper, a RELAP5-3D simulation model is set up based on the 1/16 scale experimental facility established by Texas A&M University. Also, RELAP5-3D input decks are modified to replicate the experiment procedures and the experimental results are compared with the simulation results. The results show there is a perfect match between experimental and simulation results. Radiation heat transfer dominates in the heat transfer process of high temperature gas-cooled reactor due to its high operation temperature. According to experimental research done with the RCCS facility in Texas A&M University, radiation heat transfer takes up 80% of the total heat transferred to standing pipes. In radiation heat transfer, the important parameters are view factors between surfaces. However, because of the geometrical complexity in the experimental facility, it is hard to use the numerical method or analytical view factor formula to calculate view factors. In this project, MCNP based on the Monte Carlo method is used to generate view factors for RELAP5-3D input. MCNP is powerful in setting up complicated geometry, source definition and tally application. In the end, RCCS geometry is set up using MCNP and view factors are calculated. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/151265
Book Synopsis Computer Simulation of Natural Convection in a Water-cooled Reactor Cavity Cooling System with RELAP5 by : Robert M. Scherrer
Download or read book Computer Simulation of Natural Convection in a Water-cooled Reactor Cavity Cooling System with RELAP5 written by Robert M. Scherrer and published by . This book was released on 2011 with total page 242 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis RELAP5-3D Thermal Hydraulics Computer Program Analysis Coupled with DAKOTA and STAR-CCM+ Codes by : Oscar Artajerjes Rodriguez
Download or read book RELAP5-3D Thermal Hydraulics Computer Program Analysis Coupled with DAKOTA and STAR-CCM+ Codes written by Oscar Artajerjes Rodriguez and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: kRELAP5-3D has been coupled with both DAKOTA and STAR-CCM+ in order to expand the capability of the thermal-hydraulic code and facilitate complex studies of desired systems. In the first study, RELAP5-3D was coupled with DAKOTA to perform a sensitivity study of the South Texas Project (STP) power plant during steady-state and transient scenarios. The coupled software was validated by analyzing the simulation results with respect of the physical expectations and behavior of the power plant, and thermal-hydraulic parameters which caused greatest sensitivity where identified: inlet core temperature and reactor thermal power. These variables, along with break size and discharge coefficients, were used for further investigation of the sensitivity of the RELAP5-3D LOCA transient simulation under three difference cases: two inch break, six inch break, and guillotine break. Reactor thermal power, core inlet temperature, and break size were identified as producing the greatest sensitivity; therefore, future research would include uncertainty quantification for these parameters. In the second study, a small scale experimental facility, designed to study the thermal hydraulic phenomena of the Reactor Cavity Cooling System (RCCS) for a Very High Temperature Reactor (VHTR), was used as a model to test the capabilities of coupling Star-CCM+ and RELAP5-3D. This chapter discusses the capabilities and limitations of the STAR-CCM+/RELAP5-3D coupling, and a simulation, on the RCCS facility, was performed using STAR-CCM+ to study the flow patterns where expected complex flow phenomena occur and RELAP5-3D for the complete system. The code showed inability to perform flow coupling simulations and it is unable, at this time, to handle closed loop systems. The thermal coupling simulation was successful and showed congruent qualitative results to physical expectations. The locations of large fluid vortices were located specifically in the pipes closest to the inlet of the bottom manifold. In conclusion, simulations using coupled codes were presented which greatly improved the capabilities of RELAP5-3D stand-alone and computational time required to perform complex thermal-hydraulic studies. These improvements show greatly benefit for industrial applications in order to perform large scale thermal-hydraulic systems studies with greater accuracy while minimizing simulation time. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/148410
Book Synopsis Experimental Study of the Thermal-hydraulic Phenomena in the Reactor Cavity Cooling System and Analysis of the Effects of Graphite Dispersion by : Rodolfo Vaghetto
Download or read book Experimental Study of the Thermal-hydraulic Phenomena in the Reactor Cavity Cooling System and Analysis of the Effects of Graphite Dispersion written by Rodolfo Vaghetto and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal hydraulic phenomena in a Reactor Cavity Cooling System (RCCS). The small scale RCCS experimental facility (16.5cm x 16.5cm x 30.4cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it in the environment by mixing with cold water in a large tank. PIV technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel and pipes walls and air. 10g of a fine graphite powder (particle size average 2 [mu]m) were injected into the cavity through a spraying nozzle placed at the bottom of the vessel. Temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces which was related to an increase in their emissivity. The results contribute to the understanding of the RCCS capability in case of an accident scenario.
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 Experimental Investigations in a Reactor Cavity Cooling System with Advanced Instrumentation for the Study of Instabilities, Oscillations, and Transients by : Casey Allen Tompkins
Download or read book Experimental Investigations in a Reactor Cavity Cooling System with Advanced Instrumentation for the Study of Instabilities, Oscillations, and Transients written by Casey Allen Tompkins and published by . This book was released on 2017 with total page 176 pages. Available in PDF, EPUB and Kindle. Book excerpt: A research team at University of Wisconsin - Madison designed and constructed a 1/4 height scaled experimental facility to study two-phase natural circulation cooling in a water-based reactor cavity cooling system (WRCCS) for decay heat removal in an advanced high temperature reactor. The facility is capable of natural circulation operation scaled for simulated decay heat removal (up to 28.5kWm−2 (45kW) input power, which is equivalent to 14.25kWm−2 (6.8MW) at full scale) and pressurized up to 2 bar. The UW-WRCCS facility has been used to study instabilities and oscillations observed during natural circulation flow due to evaporation of the water inventory. During two-phase operation, the system exhibits flow oscillations and excursions, which cause thermal oscillations in the structure. This can cause degradation in the mechanical structure at welds and limit heat transfer to the coolant. The facility is equipped with wire mesh sensors (WMS) that enable high-resolution measurements of the void fraction and steam velocities in order to study the instability's and oscillation's growth and decay during transient operation. Multiple perturbations to the system's operating point in pressure and inlet throttling have shown that the oscillatory behavior present under normal two-phase operating conditions can be damped and removed. Furthermore, with steady-state modeling it was discovered that a flow regime transition instability is the primary cause of oscillations in the UW-WRCCS facility under unperturbed conditions and that proper orifice selection can move the system into a stable operating regime.
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 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 Relap5-3d Model Validation and Benchmark Exercises for Advanced Gas Cooled Reactor Application by : Eugene James Thomas Moore
Download or read book Relap5-3d Model Validation and Benchmark Exercises for Advanced Gas Cooled Reactor Application written by Eugene James Thomas Moore and published by . This book was released on 2006 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: High-temperature gas-cooled reactors (HTGR) are passively safe, efficient, and economical solutions to the world's energy crisis. HTGRs are capable of generating high temperatures during normal operation, introducing design challenges related to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential to ensure the adequacy of safety features, such as the reactor cavity cooling system (RCCS). Modeling abilities of system analysis codes, used to develop an understanding of light water reactor phenomenology, need to be proven for HTGRs. RELAP5-3D v2.3.6 is used to generate two reactor plant models for a code-to-code and a code-to-experiment benchmark problem. The code-to-code benchmark problem models the Russian VGM reactor for pressurized and depressurized pressure vessel conditions. Temperature profiles corresponding to each condition are assigned to the pressure vessel heat structure. Experiment objectives are to calculate total thermal energy transferred to the RCCS for both cases. Qualitatively, RELAP5-3D's predictions agree closely with those of other system codes such as MORECA and Thermix. RELAP5-3D predicts that 80% of thermal energy transferred to the RCCS is radiant. Quantitatively, RELAP5-3D computes slightly higher radiant and convective heat transfer rates than other system analysis codes. Differences in convective heat transfer rate arise from the type and usage of convection models. Differences in radiant heat transfer stem from the calculation of radiation shape factors, also known as view or configuration factors. A MATLAB script employs a set of radiation shape factor correlations and applies them to the RELAP5-3D model. This same script is used to generate radiation shape factors for the code-to-experiment benchmark problem, which uses the Japanese HTTR reactor to determine temperature along the outside of the pressure vessel. Despite lacking information on material properties, emissivities, and initial conditions, RELAP5-3D temperature trend predictions closely match those of other system codes. Compared to experimental measurements, however, RELAP5-3D cannot capture fluid behavior above the pressure vessel. While qualitatively agreeing over the pressure vessel body, RELAP5-3D predictions diverge from experimental measurements elsewhere. This difference reflects the limitations of using a system analysis code where computational fluid dynamics codes are better suited.
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 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
