1D and 3D Simulation of a Water-cooled Reactor Cavity Cooling System Experimental Facility PDF Download
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Author:
Publisher:
ISBN:
Category : Electronic dissertations
Languages : en
Pages : 323
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Book Description
Natural circulation -- Reactor cavity cooling system -- 1D systems CFD -- 3D CFD -- 1D/3D coupling -- Natuurlike sirkulasie -- Reaktor ruimte verkoeling stelsel -- 1D stelsel berekenings vloei dinamika -- 3D berekenings vloei dinamika -- 1D/3D koppeling.
Author:
Publisher:
ISBN:
Category : Electronic dissertations
Languages : en
Pages : 323
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Book Description
Natural circulation -- Reactor cavity cooling system -- 1D systems CFD -- 3D CFD -- 1D/3D coupling -- Natuurlike sirkulasie -- Reaktor ruimte verkoeling stelsel -- 1D stelsel berekenings vloei dinamika -- 3D berekenings vloei dinamika -- 1D/3D koppeling.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 320
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Book Description
The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) 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 stead-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 experimental data produced during the stead-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.
Author:
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ISBN:
Category :
Languages : en
Pages :
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Book Description
The objective of this project is to study the fundamental physical phenomena in the reactor cavity cooling system (RCCS) of very high-temperature reactors (VHTRs). One of the primary design objectives is to assure that RCCS acts as an ultimate heat sink capable of maintaining thermal integrity of the fuel, vessel, and equipment within the reactor cavity for the entire spectrum of postulated accident scenarios. Since construction of full-scale experimental test facilities to study these phenomena is impractical, it is logical to expect that computational fluid dynamics (CFD) simulations will play a key role in the RCCS design process. An important question then arises: To what extent are conventional CFD codes able to accurately capture the most important flow phenomena, and how can they be modified to improve their quantitative predictions? Researchers are working to tackle this problem in two ways. First, in the experimental phase, the research team plans to design and construct an innovative platform that will provide a standard test setting for validating CFD codes proposed for the RCCS design. This capability will significantly advance the state of knowledge in both liquid-cooled and gas-cooled (e.g., sodium fast reactor) reactor technology. This work will also extend flow measurements to micro-scale levels not obtainable in large-scale test facilities, thereby revealing previously undetectable phenomena that will complement the existing infrastructure. Second, in the computational phase of this work, numerical simulation of the flow and temperature profiles will be performed using advanced turbulence models to simulate the complex conditions of flows in critical zones of the cavity. These models will be validated and verified so that they can be implemented into commercially available CFD codes. Ultimately, the results of these validation studies can then be used to enable a more accurate design and safety evaluation of systems in actual nuclear power applications (both during normal operation and accident scenarios).
Author: Jae Jong Oh
Publisher:
ISBN:
Category :
Languages : en
Pages : 188
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Book Description
Author: Kurt D. Hamman
Publisher:
ISBN: 9781339321523
Category : Nuclear reactors
Languages : en
Pages : 412
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Book Description
Current research in advanced reactor designs has focused on passive safety systems, where in the event of a loss of cooling to the reactor core, excess heat will be removed by a passive safety heat removal system. A safety system is classified as 'passive' because it does not require a pump to circulate the fluid (i.e., forced circulation) or operator action to maintain cooling. The system relies on the natural circulation of a fluid (i.e., fluid density differences and gravity) to transfer the heat. Passive safety system designs include features that enhance natural circulation, such as using smooth pipes, minimizing flow obstructions, and maximizing density differences, which increase fluid velocity and hence the removal of more heat. This research consisted of a CFD study of wall-bounded transitional flows and a passive reactor cavity cooling system. Yet in an effort to better understand fundamental phenomena, relative to the limits of natural circulation turbulence modeling, only forced circulation CFD analyses were performed. The initial phase of this research consisted of two types of CFD studies: 2D entropy generation rate boundary layer analyses of an isothermal transitional fluid flow over a flat plate, and 3D thermal performance analyses of a 1/4-scale experimental air reactor cavity cooling system. The 2D flat plate boundary layer studies were important in that they provided insight into flow features, such as boundary layer development and entropy generation rate, in the 3D RCCS ducts as the air transitions from laminar to turbulent flow. Using the results of the initial study as a baseline, this work analyzed the viscous and thermal boundary layer development, including estimating the entropy generation rate, in the heated duct section of the RCCS, which is characterized by nonuniform flow and heat transfer. A new engineering design process was developed, which incorporates not only traditional heat transfer and fluid flow (HTFF) analysis techniques but entropy generation minimization (EGM) concepts as well. This analysis process was successfully applied to the existing 1/4-scale experimental air RCCS, resulting in the identification of the primary entropy dissipation mechanism and an improved design.
Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201274106
Category : Science
Languages : en
Pages : 423
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Book Description
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.
Author: N. B. Siccama
Publisher:
ISBN:
Category :
Languages : en
Pages : 45
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Book Description
Author: K. G. Heal
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Book Description
Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201253101
Category : Business & Economics
Languages : en
Pages : 639
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Book Description
This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.
Author: Francesco D'Auria
Publisher: Woodhead Publishing
ISBN: 0081006799
Category : Technology & Engineering
Languages : en
Pages : 1200
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Book Description
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
