Download or read book Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

Download or read book Analytical and Experimental Study of The Effects of Non Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor written by and published by . This book was released on 2003 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems.

Download or read book Experimental and Analytical Study of the Effects of Noncondensable Gas in a Passive Condenser System written by Seungmin Oh and published by . This book was released on 2004 with total page 290 pages. Available in PDF, EPUB and Kindle. Book excerpt: Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR) is a passive condenser system which is designed to remove energy from the reactor containment during a postulated reactor accident. The presence of noncondensable gas in the vapor can greatly reduce the performance of condensers. Hence a detailed knowledge of the heat removal performance of the PCCS in the presence of noncondensable gas is crucial for the safety and design optimization of the SBWR. The purpose of the present study is the experimental and theoretical investigation of the effects of noncondensable gas in a passive condenser system. Condensation experiments were performed for a vertical tube submerged in water pool. The present experimental data provide a new database for complete condensation, cyclic venting and through flow modes of the passive condenser. Cyclic venting mode was simulated by a control volume analysis Analysis results showed that venting period decreases with noncondensable gas fraction. It was found that inception of venting can occur before the condenser is fully filled with noncondensable gas. A boundary layer model was developed for the prediction of the film condensation with noncondensable gas in a vertical tube. Full set of the governing equations for the liquid film and vapor-gas mixture regions were solved. A heat and mass analogy model was also developed with a specific purpose for use in the thermal hydraulic system analysis code. In the vapor-gas mixture region, general momentum, heat and mass transport relations derived by analytic method were used with the consideration of surface suction effect. The predictions from the models were compared with the experimental data and the agreement was satisfactory. A mechanistic condensation correlation was developed based on the experimental data and the analysis results. It contains all the heat transfer components in its functional relationships. New correlation can provide accurate estimation of local condensation heat transfer coefficient for wide range of operating parameters. The assessment of wall condensation models in RELAP5 code was performed. Experimental conditions were simulated with RELAP5. Code simulation showed quite different results compared with data. Therefore, the condensation model in RELAP5 needs to be improved.

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

Download or read book Proceedings of the International Topical Meeting on Advanced Reactors Safety written by and published by . This book was released on 1997 with total page 746 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Advanced Methodology to Simulate Boiling Water Reactor Transient Using Coupled Thermal hydraulic neutron kinetic Codes written by and published by . This book was released on 2016 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Super Light Water Reactors and Super Fast Reactors written by Yoshiaki Oka and published by Springer Science & Business Media. This book was released on 2010-07-01 with total page 664 pages. Available in PDF, EPUB and Kindle. Book excerpt: Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain the understanding of the conceptual design elements and specific analysis methods of supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference to engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology.

Download or read book Frequency response Analysis of Steam Voids to Sinusoidal Power Modulation in a Thin walled Boiling Water Coolant Channel written by Carl C. St. Pierre and published by . This book was released on 1965 with total page 150 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Experimental study on thermal hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR written by Takashi Satoh and published by . This book was released on 1998 with total page 77 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Effects of Condensation Modeling on Transient Behavior of Pressurized Water Reactors written by and published by . This book was released on 1982 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In simulating pressurized water reactor (PWR) transients with large-scale systems codes such as TRAC and RELAP, the effect of condensation has been recognized as a controlling mechanism in the prediction of plant response. For transients involving contraction of or loss of primary coolant, the rate of condensation (primarily in the pressurizer) controls the system refill characteristics. Several separate but interacting phenomena occur during the process of pressurizer refill: steam compression, system heat losses, thermal stratification or mixing of liquid, and condensation. The relative importance of each of these processes and the degree of interaction between them during different transients is very complex. The existing condensation models do not adequately describe the interplay between these effects and this leads to uncertainties in the predicted system response. Further experimental data and code assessment are required to provide data necessary for improving condensation models. Three examples of transients involving uncertainties introduced by condensation modeling are (1) pressurized thermal shock (PTS) transients, (2) small break loss-of-coolant accidents (SBLOCA), and (3) steam generator tube ruptures (SGTR).

Download or read book A Thermal hydraulic Code for Transient Analysis in a Channel with a Rod Bundle written by and published by . This book was released on 1995 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

Download or read book Transient Heat Transfer in Reactor Coolant Channels written by Ralph P. Stein and published by . This book was released on 1957 with total page 46 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book An Advanced Frequency domain Code for Boiling Water Reactor BWR Stability Analysis and Design written by Martin Schraner and published by . This book was released on 2008 with total page 169 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Current Programs written by and published by . This book was released on 1977 with total page 570 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book INIS Atomindex written by and published by . This book was released on 1995 with total page 640 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Assessment of MIT and UCB Wall Condensation Tests and of the Pre release RELAP5 written by and published by . This book was released on 1995 with total page 96 pages. Available in PDF, EPUB and Kindle. Book excerpt: In recent years, a new class of reactor designs has been proposed that utilize passive safety systems. General Electric has developed a Simplified Boiling Water Reactor (SBWR) design that relies on such passive systems. The SBWR has two passive cooling systems that involve energy transfer by condensation. These are the isolation condenser system (ICS) and the passive containment cooling systems (PCCS). It is important that such heat transfer phenomena be correctly understood and quantified. The General Electric Company has sponsored tests at the Massachusetts Institute of Technology (MIT) and at the University of California at Berkeley (UCB) to obtain data simulating PCCS conditions. Data was obtained with pure steam, steam-air mixtures and steam-helium mixtures. INEL has been contracted by the NRC to evaluate these tests and assess existing condensation heat transfer correlations against the test data. This report assesses the relevance of the tests to SBWR conditions and shows RELAP5/MOD3.2 predictions of the tests.