Read Books Online and Download eBooks, EPub, PDF, Mobi, Kindle, Text Full Free.
Integrated Fuel Performance And Thermal Hydraulic Sub Channel Models For Analysis Of Sodium Fast Reactors
Download Integrated Fuel Performance And Thermal Hydraulic Sub Channel Models For Analysis Of Sodium Fast Reactors full books in PDF, epub, and Kindle. Read online Integrated Fuel Performance And Thermal Hydraulic Sub Channel Models For Analysis Of Sodium Fast Reactors ebook anywhere anytime directly on your device. Fast Download speed and no annoying ads. We cannot guarantee that every ebooks is available!
Book Synopsis Integrated Fuel Performance and Thermal-hydraulic Sub-channel Models for Analysis of Sodium Fast Reactors by : Joseph William Fricano
Download or read book Integrated Fuel Performance and Thermal-hydraulic Sub-channel Models for Analysis of Sodium Fast Reactors written by Joseph William Fricano and published by . This book was released on 2012 with total page 257 pages. Available in PDF, EPUB and Kindle. Book excerpt: Sodium Fast Reactors (SFR) show promise as an effective way to produce clean safe nuclear power while properly managing the fuel cycle. Accurate computer modeling is an important step in the design and eventual licensing of SFRs. The objective of this work was to couple a model for metal fuel performance to a sub-channel analysis code to more precisely predict critical phenomena that could lead to pin failure for steady-state and transient scenarios. The fuel code that was used is the recently developed and benchmarked FEAST-METAL code. The sub-channel analysis code that was selected is COBRA-IV-I. This code was updated with current correlations for sodium for pressure drop, mixing, and heat transfer. The new code, COBRA-IV-I-MIT was then validated with experimental data from the Oak Ridge National Laboratory (ORNL) 19-Pin Bundle, the Toshiba 37-Pin Bundle, and the Westinghouse Advanced Reactors Division (WARD) 61-Pin Bundle. Important topics that were addressed for coupling the codes include the following. The importance of azimuthal effects in the fuel pin: FEAST only evaluates the fuel in two-dimensions, assuming azimuthal symmetry; however, coupling to COBRA produces an azimuthal temperature distribution. The acceptability of assuming a two-dimensional fuel rod with an average temperature was examined. Furthermore, how the fuel pin evolves over time affects the assembly geometry. How well a two-dimensional fuel rod allows for an accurate description of the changing assembly geometry was also considered. Related to this was how the evolution of the assembly geometry affects its thermal hydraulic behavior, which determined the exact form of coupling between the codes. Ultimately one-way coupling was selected with azimuthal temperature averaging around the fuel pin. The codes were coupled using a wrapper, the COBRA And FEAST Executer (CAFE), written in the Python programming language. Data from EBR-II was used to confirm and verify CAFE. It was found that the number of axial nodes used in FEAST can have a large effect on the result. Finally FEAST was used to parametrically study three different pin designs: driver fuel, radial blanket, and tight pitch breed and bum fuel. This study provides data for pin expected life in assembly design.
Book Synopsis Thermal-hydraulic Analysis of Innovative Fuel Configurations for the Sodium Fast Reactor by : Matthew J. Memmott
Download or read book Thermal-hydraulic Analysis of Innovative Fuel Configurations for the Sodium Fast Reactor written by Matthew J. Memmott and published by . This book was released on 2009 with total page 458 pages. Available in PDF, EPUB and Kindle. Book excerpt: (Cont.) As an application of this subchannel model, duct ribs were explored as a means of reducing core outlet temperature peaking within the fuel assemblies. The performance of the annular and bottle-shaped fuel was also investigated using this subchannel model. The annular fuel configurations are best suited for low conversion ratio cores. The magnitude of the power uprate enabled by metal annular fuel in the CR = 0.25 cores is 20%, and is limited by the FCCI constraint during a hypothetical flow blockage of the inner-annular channel due to the small diameters of the inner-annular flow channel (3.6 mm). On the other hand, a complete blockage of the hottest inner-annular flow channel in the oxide fuel case results in sodium boiling, which renders the annular oxide fuel concept unacceptable for use in a SFR. The bottle-shaped fuel configurations are best suited for high conversion ratio cores. In the CR = 0.71 cores, the bottle-shaped fuel configuration reduces the overall core pressure drop in the fuel channels by up to 36.3%. The corresponding increase in core height with bottle-shaped fuel is between 15.6% and 18.3%. A full-plant RELAP5-3D model was created to evaluate the transient performance of the base and innovative fuel configurations during station blackout and UTOP transients. The transient analysis confirmed the good thermal-hydraulic performance of the annular and bottle-shaped fuel designs with respect to their respective solid fuel pin cases.
Book Synopsis Dynamic Simulation of Sodium Cooled Fast Reactors by : G Vaidyanathan
Download or read book Dynamic Simulation of Sodium Cooled Fast Reactors written by G Vaidyanathan and published by CRC Press. This book was released on 2022-11-18 with total page 289 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book provides the basis of simulating a nuclear plant, in understanding the knowledge of how such simulations help in assuring the safety of the plants, thereby protecting the public from accidents. It provides the reader with an in-depth knowledge about modeling the thermal and flow processes in a fast reactor and gives an idea about the different numerical solution methods. The text highlights the application of the simulation to typical sodium-cooled fast reactor. The book • Discusses mathematical modeling of the heat transfer process in a fast reactor cooled by sodium. • Compares different numerical techniques and brings out the best one for the solution of the models. • Provides a methodology of validation based on experiments. • Examines modeling and simulation aspects necessary for the safe design of a fast reactor. • Emphasizes plant dynamics aspects, which is important for relating the interaction between the components in the heat transport systems. • Discusses the application of the models to the design of a sodium-cooled fast reactor It will serve as an ideal reference text for senior undergraduate, graduate students, and academic researchers in the fields of nuclear engineering, mechanical engineering, and power cycle engineering.
Book Synopsis Thermal-hydraulic Numerical Simulation of Fuel Sub-assembly for Sodium-cooled Fast Reactor by : Aakanksha Saxena
Download or read book Thermal-hydraulic Numerical Simulation of Fuel Sub-assembly for Sodium-cooled Fast Reactor written by Aakanksha Saxena and published by . This book was released on 2014 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR).First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO_U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry.The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (~20°C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (~1-2°C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures.
Book Synopsis Multi-scale Multi-Physics Modeling of Metallic Fuel and Thermal Hydraulics of Sodium Fast Reactors in a Subchannel Approach by : Ahmed Mohamed Nabil Hassanein Aly
Download or read book Multi-scale Multi-Physics Modeling of Metallic Fuel and Thermal Hydraulics of Sodium Fast Reactors in a Subchannel Approach written by Ahmed Mohamed Nabil Hassanein Aly and published by . This book was released on 2022 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors by : Ferry Roelofs
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
Book Synopsis Thermal Hydraulic and Fuel Performance Analysis for Innovative Small Light Water Reactor Using VIPRE-01 and FRAPCON-3 by : Anh T. Mai
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.
Book Synopsis FAST REACTOR FUEL ASSEMBLY THERMAL HYDRAULIC DEVELOPMENT PROGRAM by :
Download or read book FAST REACTOR FUEL ASSEMBLY THERMAL HYDRAULIC DEVELOPMENT PROGRAM written by and published by . This book was released on 1970 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book REPP written by R. M. Hiatt and published by . This book was released on 1969 with total page 196 pages. Available in PDF, EPUB and Kindle. Book excerpt: REPP, a digital computer method for designing pressure water and boiling water reactor cores within specified heat transfer and fuel centerline temperature limits is presented. The method incorporates the Westinghouse W-2 and W-3 empirical correlations and a theoretical hot channel model to predict burnout conditions in a rod bundle. Two geometries are considered; rods in a triangular array and rods in a square lattice. The heat transfer problem solved is a one-dimensional analysis. Pressure drop is considered for four types of fuel-pin spacers. Variable heat generation rate through the fuel-pin and sintering in low density fuels are also included.
Book Synopsis Reynolds-Averaged Navier-Stokes-based Thermal-hydraulic Investigation of the Sodium Cooled Fast Reactor Standard Fuel Bundle, Altered Bundles, and Spacer Grids by : Jeffrey Gordon Smith
Download or read book Reynolds-Averaged Navier-Stokes-based Thermal-hydraulic Investigation of the Sodium Cooled Fast Reactor Standard Fuel Bundle, Altered Bundles, and Spacer Grids written by Jeffrey Gordon Smith and published by . This book was released on 2010 with total page 280 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Super Light Water Reactors and Super Fast Reactors by : Yoshiaki Oka
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.
Book Synopsis Neutronics and Thermal-hydraulics Coupling by : Maxime Guyot
Download or read book Neutronics and Thermal-hydraulics Coupling written by Maxime Guyot and published by . This book was released on 2014 with total page 239 pages. Available in PDF, EPUB and Kindle. Book excerpt: This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and recriticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios.During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. Inthe multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level.In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling.
Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1976 with total page 658 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors by : Matthew Thomas Weathered
Download or read book Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors written by Matthew Thomas Weathered and published by . This book was released on 2017 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: The thermal hydraulic properties of liquid sodium make it an attractive coolant for use in Generation IV reactors. The liquid metal's high thermal conductivity and low Prandtl number increases efficiency in heat transfer at fuel rods and heat exchangers, but can also cause features such as high magnitude temperature oscillations and gradients in the coolant. Currently, there exists a knowledge gap in the mechanisms which may create these features and their effect on mechanical structures in a sodium fast reactor. Two of these mechanisms include thermal striping and thermal stratification. Thermal striping is the oscillating temperature field created by the turbulent mixing of non-isothermal flows. Usually this occurs at the reactor core outlet or in piping junctions and can cause thermal fatigue in mechanical structures. Meanwhile, thermal stratification results from large volumes of non-isothermal sodium in a pool type reactor, usually caused by a loss of coolant flow accident. This stratification creates buoyancy driven flow transients and high temperature gradients which can also lead to thermal fatigue in reactor structures. In order to study these phenomena in sodium, a novel method for the deployment of optical fiber temperature sensors was developed. This method promotes rapid thermal response time and high spatial temperature resolution in the fluid. The thermal striping and stratification behavior in sodium may be experimentally analyzed with these sensors with greater fidelity than ever before. Thermal striping behavior at a junction of non-isothermal sodium was fully characterized with optical fibers. An experimental vessel was hydrodynamically scaled to model thermal stratification in a prototypical sodium reactor pool. Novel auxiliary applications of the optical fiber temperature sensors were developed throughout the course of this work. One such application includes local convection coefficient determination in a vessel with the corollary application of level sensing. Other applications were cross correlation velocimetry to determine bulk sodium flow rate and the characterization of coherent vortical structures in sodium with temperature frequency data. The data harvested, instrumentation developed and techniques refined in this work will help in the design of more robust reactors as well as validate computational models for licensing sodium fast reactors.
Book Synopsis Applying Uncertainty and Sensitivity on Thermal Hydraulic Subchannel Analysis for the Multi-application Small Light Water Reactor by : Adam Brigantic
Download or read book Applying Uncertainty and Sensitivity on Thermal Hydraulic Subchannel Analysis for the Multi-application Small Light Water Reactor written by Adam Brigantic and published by . This book was released on 2014 with total page 115 pages. Available in PDF, EPUB and Kindle. Book excerpt: Small modular reactors (SMRs) are a recent advancement in commercial nuclear reactor design with growing interest worldwide. New SMR concepts, such as the Multi-Application Small Light Water Reactor (MASLWR), must undergo a licensing processes established by the U.S. Nuclear Regulatory Commission (NRC) prior to commercial operation. Given the lack of historical, full scale operating experience, a general uncertainty and sensitivity analysis methodology was developed to help aid SMR designs through this process. Uncertainty was quantified through the empirical cumulative distribution function (ECDF) created from a desired data set. Linear regression techniques were applied to measure sensitivity. This methodology was demonstrated through the thermal hydraulic subchannel analysis of the MASLWR concept using RELAP5-3D Version 4.0.3 and VIPRE-01 Mod 2.2.1. Twelve uncertain input parameters were selected. System response uncertainty in the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel temperature, and maximum clad temperature was evaluated. These figures were shown to satisfy U.S. NRC regulatory requirements for steady state operation at the 95 percent probability and 95 percent confidence level under the evaluated conditions. Sensitivity studies showed input parameters affecting local power generation within the core had a large influence on MDNBR, maximum fuel temperature, and maximum clad temperature.
Book Synopsis Single- and Two-phase Flow Modeling for Coupled Neutronics/thermal-hydraulics Transient Analysis of Advanced Sodium-cooled Fast Reactors by : Aurélia Chenu
Download or read book Single- and Two-phase Flow Modeling for Coupled Neutronics/thermal-hydraulics Transient Analysis of Advanced Sodium-cooled Fast Reactors written by Aurélia Chenu and published by . This book was released on 2011 with total page 237 pages. Available in PDF, EPUB and Kindle. Book excerpt:
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:
