Download or read book Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data written by International Atomic Energy Agency and published by . This book was released on 2012 with total page 243 pages. Available in PDF, EPUB and Kindle. Book excerpt: Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses is an International Atomic Energy Agency (IAEA) activity designed to facilitate international cooperative research and promote information exchange on computer codes for thermalhydraulic safety analyses. The objective is to enhance the safety analysis capabilities of the participants and the effective use of their resources through this international cooperation. This report provides a comparison of the results obtained from eight participating organisations from six countries, utilizing four different computer codes. General conclusions are reached and recommendations made
Download or read book Intercomparison and Validation of Computer Codes for Thermohydraulic Safety Analysis of Heavy Water Reactors IAEA TECDOC Series written by International Atomic Energy Agency and published by . This book was released on 2004 with total page 130 pages. Available in PDF, EPUB and Kindle. Book excerpt: Intercomparison and validation of computer codes used in different countries for thermal hydraulics safety analysis of heavy water reactors (HWRs) enhances the confidence in the predictions made by these codes. A set of reliable experimental data is necessary for conducting such intercomparison and validation exercises. Experimental results from a large-loss of coolant accident (LOCA) test simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd. (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participat.
Download or read book Intercomparison and Validation of Computer Codes for Thermalhydraulic Safety Analysis of Heavy Water Reactors written by International Atomic Energy Agency and published by . This book was released on 2004 with total page 140 pages. Available in PDF, EPUB and Kindle. Book excerpt: Intercomparison and validation of computer codes used in different countries for thermal hydraulics safety analysis of heavy water reactors (HWRs) enhances the confidence in the predictions made by these codes. A set of reliable experimental data is necessary for conducting such intercomparison and validation exercises. Experimental results from a large-loss of coolant accident (LOCA) test simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd. (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participating countries, utilizing four different computer codes. General conclusions are reached and recommondations made.
Download or read book Comparison of Predictions from the Reactor Primary System Decompression Code RELAP3 with Decompression Data from the Semiscale Blowdown and Emergency Core Cooling ECC Project written by C. E. Slater and published by . This book was released on 1970 with total page 30 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 with total page 444 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 Comparison of the PLTEMP Code Flow Instability Predictions with Measurements Made with Electrically Heated Channels for the Advanced Test Reactor written by and published by . This book was released on 2011 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodate an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code. Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm2. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow instability for these tests leads one to the conclusion that the same method should be able to predict the onset of flow instability in the MURR reasonably well.
Download or read book Benchmarking Severe Accident Computer Codes for Heavy Water Reactor Applications written by International Atomic Energy Agency and published by . This book was released on 2014-02-19 with total page 311 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication reports on the results of an IAEA cooperated research project (CRP) on benchmarking severe accident computer codes for heavy water reactor applications. The CRP scope included the identification and selection of a severe accident sequence, selection of appropriate geometrical and boundary conditions, conduct of benchmark analyses, comparison of the results of all code outputs, evaluation of the capabilities of computer codes to predict important severe accident phenomena, and the proposal of necessary code improvements and/or new experiments to reduce uncertainties. The summary report provides a comparison of key results obtained from five participating countries and concludes with lessons learned and recommendations for the future.
Download or read book Proceedings of the Third International Topical Meeting on Reactor Thermal Hydraulics Newport Rhode Island U S A October 15 18 1985 written by Chong Chiu and published by . This book was released on 1985 with total page 684 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Thermal hydraulic Response of the Semiscale Mod 1 System written by and published by . This book was released on 1975 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Selected experimental thermal-hydraulic data from the recent isothermal blowdown test series performed in the Semiscale Mod-1 geometry are analyzed from an experimental viewpoint with emphasis on explaining differences between the data and expected results. Comparisons are made between the trends measured by the system instrumentation and the trends predicted by analytical tools, including the RELAP4 computer code, to aid in understanding the interactions between phenomena occurring in different parts of the system. The analyses presented are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system of a pressurized water reactor during a postulated loss-of-coolant accident (LOCA). 20 references. (auth).
Download or read book Government Reports Announcements Index written by and published by . This book was released on 1984 with total page 1408 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Proceedings of Second International Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations written by Jirō Wakabayashi and published by . This book was released on 1986 with total page 1270 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Transactions of the American Nuclear Society written by American Nuclear Society and published by . This book was released on 1984 with total page 1088 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Proceedings Annual Conference written by Canadian Nuclear Society. Conference and published by . This book was released on 1989 with total page 404 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 1988 with total page 794 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Regulatory and Technical Reports written by U.S. Nuclear Regulatory Commission. Division of Technical Information and Document Control and published by . This book was released on 1980 with total page 596 pages. Available in PDF, EPUB and Kindle. Book excerpt: Includes indexes.
Download or read book Energy Research Abstracts written by and published by . This book was released on 1985 with total page 900 pages. Available in PDF, EPUB and Kindle. Book excerpt: Semiannual, with semiannual and annual indexes. References to all scientific and technical literature coming from DOE, its laboratories, energy centers, and contractors. Includes all works deriving from DOE, other related government-sponsored information, and foreign nonnuclear information. Arranged under 39 categories, e.g., Biomedical sciences, basic studies; Biomedical sciences, applied studies; Health and safety; and Fusion energy. Entry gives bibliographical information and abstract. Corporate, author, subject, report number indexes.
