› › Thermal Hydraulic Analysis Of Innovative Fuel Configurations For The Sodium Fast Reactor

Thermal-hydraulic Analysis of Innovative Fuel Configurations for the Sodium Fast Reactor

Author : Matthew J. Memmott

Publisher :

Page : 458 pages

File Size : 15,63 MB

Release : 2009

Category :

ISBN :

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(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.



Nuclear Reactor Thermal Hydraulics

Author : Robert E. Masterson

Publisher : CRC Press

Page : 2352 pages

File Size : 48,6 MB

Release : 2019-08-21

Category : Science

ISBN : 1351849514

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Book Description

Nuclear Thermal-Hydraulic Systems provides a comprehensive approach to nuclear reactor thermal-hydraulics, reflecting the latest technologies, reactor designs, and safety considerations. The text makes extensive use of color images, internet links, computer graphics, and other innovative techniques to explore nuclear power plant design and operation. Key fluid mechanics, heat transfer, and nuclear engineering concepts are carefully explained, and supported with worked examples, tables, and graphics. Intended for use in one or two semester courses, the text is suitable for both undergraduate and graduate students. A complete Solutions Manual is available for professors adopting the text.



Integrated Fuel Performance and Thermal-hydraulic Sub-channel Models for Analysis of Sodium Fast Reactors

Author : Joseph William Fricano

Publisher :

Page : 257 pages

File Size : 43,78 MB

Release : 2012

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ISBN :

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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.



Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

Author : Ferry Roelofs

Publisher : Woodhead Publishing

Page : 464 pages

File Size : 11,98 MB

Release : 2018-11-30

Category : Science

ISBN : 0081019815

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Book Description

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



Thermal Hydraulic and Fuel Performance Analysis for Innovative Small Light Water Reactor Using VIPRE-01 and FRAPCON-3

Author : Anh T. Mai

Publisher :

Page : 151 pages

File Size : 22,73 MB

Release : 2012

Category : Light water reactors

ISBN :

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Book Description

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.



Thermal-Hydraulic Analysis of Nuclear Reactors

Author : Bahman Zohuri

Publisher : Springer

Page : 667 pages

File Size : 39,76 MB

Release : 2015-09-09

Category : Science

ISBN : 3319174347

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Book Description

This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play. Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental definitions of units and dimensions, thermodynamic variables, and the Laws of Thermodynamics progressing to sections on specific applications of the Brayton and Rankine cycles for power generation and projected reactor systems design issues Reinforces fundamentals of fluid dynamics and heat transfer; thermal and hydraulic analysis of nuclear reactors, two-phase flow and boiling, compressible flow, stress analysis, and energy conversion methods Includes detailed appendices that cover metric and English system units and conversions, detailed steam and gas tables, heat transfer properties, and nuclear reactor system descriptions



Nuclear Thermal Hydraulic and Two-Phase Flow

Author : Jun Wang

Publisher : Frontiers Media SA

Page : 120 pages

File Size : 41,27 MB

Release : 2018-10-11

Category :

ISBN : 2889456129

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Book Description

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.



Thermal-hydraulic Numerical Simulation of Fuel Sub-assembly for Sodium-cooled Fast Reactor

Author : Aakanksha Saxena

Publisher :

Page : 0 pages

File Size : 14,42 MB

Release : 2014

Category :

ISBN :

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Book Description

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.



Sodium-cooled Fast Reactors

Author : Masaki Morishita

Publisher : Academic Press

Page : 666 pages

File Size : 42,9 MB

Release : 2022-07-22

Category : Technology & Engineering

ISBN : 0128240776

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Book Description

Sodium-cooled Fast Reactors is the third volume in the JSME Series on Thermal and Nuclear Power Generation, which presents a comprehensive view of the latest research and activities from around the globe. Volume Editors Masaki Morishita and Hiroyuki Ohshima, along with their team of expert contributors, combine their knowledge and experience to provide a solid understanding of the history of SFRs and work carried out in Japan to date. This book uniquely includes case studies from these global regions to highlight SFR uses, benefits and challenges, focusing on their safety, design, operation, and maintenance. Unique to this publication, the JSME cover key technological advances which will shape power generation of the future, including developments in the use of AI for design. Drawing on their unique experience, the authors pass on lessons learned and best practices to support professionals and researchers in their development and design of this advanced reactor type. - Written by the leaders and pioneers in nuclear research at the Japanese Society of Mechanical Engineers and draws upon their combined wealth of knowledge and experience - Includes real examples and case studies mainly from Japan to provide a deeper learning opportunity with practical benefits - Considers the societal impact and sustainability concerns and goals throughout the discussion - Includes safety factors and considerations, as well as unique results from performance testing of SFR systems



Thermal-Hydraulics of Water Cooled Nuclear Reactors

Author : Francesco D'Auria

Publisher : Woodhead Publishing

Page : 1200 pages

File Size : 10,77 MB

Release : 2017-05-18

Category : Technology & Engineering

ISBN : 0081006799

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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