Download or read book Thermal Hydraulic Analyses of the HEU and the Proposed LEU Core Configurations of the UMass Lowell Research Reactor written by A. Amarnath and published by . This book was released on 1993 with total page 240 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book LEU HEU Mixed Core Conversion Thermal hydraulic Analysis and Coolant System Upgrade Assessment for the MIT Research Reactor written by Yinjie Zhao and published by . This book was released on 2022 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MIT Research Reactor (MITR) is in the process of converting from the current 93%-enriched U-235 highly-enriched uranium (HEU) fuel to the low enriched uranium (LEU, 20%-enriched U-235) fuel, as part of the global non-proliferation initiatives. A high-density, monolithic uraniummolybdenum (U-10Mo) fuel matrix is chosen. The fuel element design is changed from 15-plate finned HEU fuel to 19-plate unfinned LEU fuel with the same geometry. The reactor power increases from 6.0 MW to 7.0 MW thermal, and primary coolant flow rate increases from 2000 gpm to 2400 gpm. Detailed analyses were completed for initial LEU core with 22 fuel elements, and demonstrated both neutronic and thermal hydraulic safety requirements are met throughout equilibrium cycles. An alternative conversion strategy is proposed which involves a gradual transition from an all-HEU core to an all-LEU core by replacing 3 HEU fuel elements with fresh LEU fuel elements during each fuel cycle. The objectives of this study are to demonstrate that the primary coolant system can be safely modified for 2400 gpm operation, and to perform steady-state and loss-of-flow (LOF) transient thermal-hydraulic analyses for the MITR HEU-LEU transitional mixed cores to evaluate this alternative conversion strategy. The primary technical challenge for the 20% increase in primary flow rate with existing piping system is flow-induced vibration. Several experiments were performed to measure and quantify vibration acceleration and velocity on three main hydraulic components to determine if higher flowrates cause excessive vibration. The test results show that the maximum vibration velocity is 9.70 mm/s, the maximum vibration acceleration is 0.98 G at the current flow rate 2000 gpm and no significant spectral change in the vibration profile at 2550 gpm. Therefore, it can be concluded that the existing piping system can safely support 2400 gpm primary flow operation. Thermal hydraulics analysis was performed using RELAP5 MOD3.3 code and STAT7 code. The MITR transitional mixed core input models were constructed to simulate the reactor primary system. Two scenarios, steady-state and loss-of-flow transient were simulated at power level of 6 MW. RELAP5 results show that during steady state, there is significant safety margin ( 10 °C) to onset of nucleate boiling for both HEU and LEU fuel. The maximum core temperature occurs at HEU fuel in Mix-core 3, the maximum wall temperature reached was 89 °C. During the LOF transient case, the result shows that The HEU fuel element is more limiting than the LEU in transitional cores. Nucleate boiling is predicted to occur only in the HEU hot channel during the first 50 seconds after the pump coastdown. The peak cladding temperatures are much lower than the fuel temperature safety limit of UAl[subscript x] fuel plates, which is 450 °C. From the STAT7 calculation results, the operational limiting power at which onset of nucleate boiling (ONB) occurs in all cases show significant margins from the Limiting System Safety Setting (LSSS) over-power level. The lowest margin for LEU element during the mixed core transition is at Mix-7, 11.43 MW with a 4.03 MW power margin. For the HEU element, the lowest margin during the transition is at Mix-2, 8.51 MW with a 1.11 MW power margin. The location at which ONB is always expected to occur is F-Plate Stripe 1 and 4 for the LEU fuel element; side plate for the HEU fuel element with the HEU element is always more limiting.

Download or read book Masters Theses in the Pure and Applied Sciences written by Wade H. Shafer and published by Springer Science & Business Media. This book was released on 2012-12-06 with total page 426 pages. Available in PDF, EPUB and Kindle. Book excerpt: Masters Theses in the Pure and Applied Sciences was first conceived, published, and disseminated by the Center for Information and Numerical Data Analysis and Synthesis (CINDAS)* at Purdue University in 1957, starting its coverage of theses with the academic year 1955. Beginning with Volume 13, the printing and dis semination phases of the activity were transferred to University Microfilms/Xerox of Ann Arbor, Michigan, with the thought that such an arrangement would be more beneficial to the academic and general scientific and technical community. After five years of this joint undertaking we had concluded that it was in the interest of all concerned if the printing and distribution of the volumes were handled by an international publishing house to assure improved service and broader dissemination. Hence, starting with Volume 18, Masters Theses in the Pure and Applied Sciences has been disseminated on a worldwide basis by Plenum Publishing Corporation of New York, and in the same year the coverage was broadened to include Canadian universities. All back issues can also be ordered from Plenum. We have reported in Volume 38 (thesis year 1993) a total of 13,787 thesis titles from 22 Canadian and 164 United States universities. We are sure that this broader base for these titles reported will greatly enhance the value of this impor tant annual reference work. While Volume 38 reports theses submitted in 1993, on occasion, certain uni versities do report theses submitted in previous years but not reported at the time.

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:

Download or read book Masters Theses in the Pure and Applied Sciences written by Sade H Shafer and published by Springer Science & Business Media. This book was released on 1995 with total page 440 pages. Available in PDF, EPUB and Kindle. Book excerpt: Cited in Sheehy, Chen, and Hurt . Volume 38 (thesis year 1993) reports a total of 13,787 thesis titles from 22 Canadian and 164 US universities. As in previous volumes, thesis titles are arranged by discipline and by university within each discipline. Any accredited university or college with a grad

Download or read book Thermal Hydraulic Limits Analysis for the Massachusetts Institute of Technology Research Reactor Low Enrichment Uranium Core Conversion Using Statistical Propagation of Parametric Uncertainties written by Keng-Yen Chiang and published by . This book was released on 2012 with total page 171 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor power upgraded from 6 MW to 7 MW is proposed in order to maintain the same reactor performance of the HEU core. Previous approaches in analyzing the impact of engineering uncertainties on thermal hydraulic limits via the use of engineering hot channel factors (EHCFs) were unable to explicitly quantify the uncertainty and confidence level in reactor parameters. The objective of this study is to develop a methodology for MITR thermal hydraulic limits analysis by statistically combining engineering uncertainties in order to eliminate unnecessary conservatism inherent in traditional analyses. This methodology was employed to analyze the Limiting Safety System Settings (LSSS) for the MITR LEU core, based on the criterion of onset of nucleate boiling (ONB). Key parameters, such as coolant channel tolerances and heat transfer coefficients, were considered as normal distributions using Oracle Crystal Ball for the LSSS evaluation. The LSSS power is determined with 99.7% confidence level. The LSSS power calculated using this new methodology is 9.1 MW, based on core outlet coolant temperature of 60 'C, and primary coolant flow rate of 1800 gpm, compared to 8.3 MW obtained from the analytical method using the EHCFs with same operating conditions. The same methodology was also used to calculate the safety limit (SL) to ensure that adequate safety margin exists between LSSS and SL. The criterion used to calculate SL is the onset of flow instability. The calculated SL is 10.6 MW, which is 1.5 MW higher than LSSS, permitting sufficient margin between LSSS and SL.

Download or read book Heat transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor written by Liaquat Ali Khan and published by . This book was released on 1987 with total page 61 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 Thermal Hydraulic Analysis of the Oregon State TRIGA Reactor Using RELAP5 3D written by Wade R. Marcum and published by . This book was released on 2008 with total page 302 pages. Available in PDF, EPUB and Kindle. Book excerpt: Oregon State University has recently conducted a complete core conversion analysis as part of the Reduced Enrichment for Research and Test Reactors Pprogram. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the Highly Enriched Uranium (HEU) and Low Enriched Uranium (LEU) cores; for steady state and pulsed operation, calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as critical heat flux ratio (CHFR) and temperature profiles in hot channel for both the HEU and LEU cores; finally, perform accident analyses for the accident scenarios identified in the Oregon State TRIGA® Reactor (OSTR) Safety Analysis Report (SAR). RELAP5-3D Version 2.4.2 was used for all computational modeling during the thermal hydraulics analysis. This is a lumped parameter code forcing engineering assumptions to be made during the analysis. A single hot channel model's results are compared to that produced from more refined two and eight channel models in order to identify variations in thermal hydraulic characteristics as a function of spatial refinement.

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.

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

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

Download or read book Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors written by Ferry Roelofs and published by Woodhead Publishing. This book was released on 2018-11-30 with total page 464 pages. Available in PDF, EPUB and Kindle. Book excerpt: Thermal Hydraulics Aspects of Liquid Metal cooled Nuclear Reactors is a comprehensive collection of liquid metal thermal hydraulics research and development for nuclear liquid metal reactor applications. A deliverable of the SESAME H2020 project, this book is written by top European experts who discuss topics of note that are supplemented by an international contribution from U.S. partners within the framework of the NEAMS program under the U.S. DOE. This book is a convenient source for students, professionals and academics interested in liquid metal thermal hydraulics in nuclear applications. In addition, it will also help newcomers become familiar with current techniques and knowledge. - Presents the latest information on one of the deliverables of the SESAME H2020 project - Provides an overview on the design and history of liquid metal cooled fast reactors worldwide - Describes the challenges in thermal hydraulics related to the design and safety analysis of liquid metal cooled fast reactors - Includes the codes, methods, correlations, guidelines and limitations for liquid metal fast reactor thermal hydraulic simulations clearly - Discusses state-of-the-art, multi-scale techniques for liquid metal fast reactor thermal hydraulics applications

Download or read book Status and subjects of thermal hydraulic analysis for next generation LWRs with passive safety systems written by Nihon Genshiryoku Kenkyūjo. Subcommittee on Improvement of Reactor Thermal-Hydraulic Analysis Codes and published by . This book was released on 1998 with total page 194 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book 2D Thermal Hydraulic Analysis and Benchmark in Support of HFIR LEU Conversion Using COMSOL written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The research documented herein was funded by a research contract between the Research Reactors Division (RRD) of Oak Ridge National Laboratory (ORNL) and the University of Tennessee, Knoxville (UTK) Mechanical, Aerospace and Biomedical Engineering Department (MABE). The research was governed by a statement of work (SOW) which clearly defines nine specific tasks. This report is outlined to follow and document the results of each of these nine specific tasks. The primary goal of this phase of the research is to demonstrate, through verification and validation methods, that COMSOL is a viable simulation tool for thermal-hydraulic modeling of the High Flux Isotope Reactor (HFIR) core. A secondary goal of this two-dimensional phase of the research is to establish methodology and data base libraries that are also needed in the full three-dimensional COMSOL simulation to follow. COMSOL version 3.5a was used for all of the models presented throughout this report.

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

Download or read book Thermal Hydraulic and Fuel Performance Analysis for Innovative Small Light Water Reactor Using VIPRE 01 and FRAPCON 3 written by Anh T. Mai and published by . This book was released on 2012 with total page 151 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Multi-Application Small Light Water Reactor (MASLWR) is a small natural circulation pressurized light water reactor design that was developed by Oregon State University (OSU) and Idaho National Engineering and Environmental Laboratory (INEEL) under the Nuclear Energy Research Initiative (NERI) program to address the growing demand for energy and electricity. The MASLWR design is geared toward providing electricity to small communities in remote locations in developing countries where constructions of large nuclear power plants are not economical. The MASLWR reactor is designed to operate for five years without refueling and with fuel enrichment up to 8 %. In 2003, an experimental thermal hydraulic research facility also known as the OSU MASLWR Test Facility was constructed at Oregon State University to examined the performance of new reactor design and natural circulation reactor design concepts. This thesis is focused on the thermal hydraulics analysis and fuel performance analysis of the MASLWR prototypical cores with fuel enrichment of 4.25 % and 8 %. The goals of the thermal hydraulic analyses were to calculate the departure nucleate boiling ratio (DNBR) values, coolant temperature, cladding temperature and fuel temperature profiles in the hot channel of the reactor cores. The thermal hydraulic analysis was performed for steady state operation of the MASLWR prototypical cores. VIPRE Version 01 is the code used for all the computational modeling of the prototypical cores during thermal hydraulic analysis. The hot channel and hot rod results are compared with thermal design limits to determine the feasibility of the prototypical cores. The second level of analysis was performed with a fuel performance code FRAPCON for the limiting MASLWR fuel rods identified by the neutronic and thermal hydraulic analyses. The goals of the fuel performance analyses were to calculate the oxide thickness on the cladding and fission gas release (FGR). The oxide thickness results are compared with the acceptable design limits for standard fuel rods. The results in this research can be helpful for future core designs of small light water reactors with natural circulation.