RELAP5-3D Thermal Hydraulics Computer Program Analysis Coupled with DAKOTA and STAR-CCM+ Codes PDF Download
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Author: Oscar Artajerjes Rodriguez
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Category :
Languages : en
Pages :
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kRELAP5-3D has been coupled with both DAKOTA and STAR-CCM+ in order to expand the capability of the thermal-hydraulic code and facilitate complex studies of desired systems. In the first study, RELAP5-3D was coupled with DAKOTA to perform a sensitivity study of the South Texas Project (STP) power plant during steady-state and transient scenarios. The coupled software was validated by analyzing the simulation results with respect of the physical expectations and behavior of the power plant, and thermal-hydraulic parameters which caused greatest sensitivity where identified: inlet core temperature and reactor thermal power. These variables, along with break size and discharge coefficients, were used for further investigation of the sensitivity of the RELAP5-3D LOCA transient simulation under three difference cases: two inch break, six inch break, and guillotine break. Reactor thermal power, core inlet temperature, and break size were identified as producing the greatest sensitivity; therefore, future research would include uncertainty quantification for these parameters. In the second study, a small scale experimental facility, designed to study the thermal hydraulic phenomena of the Reactor Cavity Cooling System (RCCS) for a Very High Temperature Reactor (VHTR), was used as a model to test the capabilities of coupling Star-CCM+ and RELAP5-3D. This chapter discusses the capabilities and limitations of the STAR-CCM+/RELAP5-3D coupling, and a simulation, on the RCCS facility, was performed using STAR-CCM+ to study the flow patterns where expected complex flow phenomena occur and RELAP5-3D for the complete system. The code showed inability to perform flow coupling simulations and it is unable, at this time, to handle closed loop systems. The thermal coupling simulation was successful and showed congruent qualitative results to physical expectations. The locations of large fluid vortices were located specifically in the pipes closest to the inlet of the bottom manifold. In conclusion, simulations using coupled codes were presented which greatly improved the capabilities of RELAP5-3D stand-alone and computational time required to perform complex thermal-hydraulic studies. These improvements show greatly benefit for industrial applications in order to perform large scale thermal-hydraulic systems studies with greater accuracy while minimizing simulation time. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/148410
Author: Oscar Artajerjes Rodriguez
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
kRELAP5-3D has been coupled with both DAKOTA and STAR-CCM+ in order to expand the capability of the thermal-hydraulic code and facilitate complex studies of desired systems. In the first study, RELAP5-3D was coupled with DAKOTA to perform a sensitivity study of the South Texas Project (STP) power plant during steady-state and transient scenarios. The coupled software was validated by analyzing the simulation results with respect of the physical expectations and behavior of the power plant, and thermal-hydraulic parameters which caused greatest sensitivity where identified: inlet core temperature and reactor thermal power. These variables, along with break size and discharge coefficients, were used for further investigation of the sensitivity of the RELAP5-3D LOCA transient simulation under three difference cases: two inch break, six inch break, and guillotine break. Reactor thermal power, core inlet temperature, and break size were identified as producing the greatest sensitivity; therefore, future research would include uncertainty quantification for these parameters. In the second study, a small scale experimental facility, designed to study the thermal hydraulic phenomena of the Reactor Cavity Cooling System (RCCS) for a Very High Temperature Reactor (VHTR), was used as a model to test the capabilities of coupling Star-CCM+ and RELAP5-3D. This chapter discusses the capabilities and limitations of the STAR-CCM+/RELAP5-3D coupling, and a simulation, on the RCCS facility, was performed using STAR-CCM+ to study the flow patterns where expected complex flow phenomena occur and RELAP5-3D for the complete system. The code showed inability to perform flow coupling simulations and it is unable, at this time, to handle closed loop systems. The thermal coupling simulation was successful and showed congruent qualitative results to physical expectations. The locations of large fluid vortices were located specifically in the pipes closest to the inlet of the bottom manifold. In conclusion, simulations using coupled codes were presented which greatly improved the capabilities of RELAP5-3D stand-alone and computational time required to perform complex thermal-hydraulic studies. These improvements show greatly benefit for industrial applications in order to perform large scale thermal-hydraulic systems studies with greater accuracy while minimizing simulation time. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/148410
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Languages : en
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RELAP5-3D is widely used by the nuclear community to simulate general thermal hydraulic systems and has proven to be so versatile that the spectrum of transient two-phase problems that can be analyzed has increased substantially over time. To accommodate the many new types of problems that are analyzed by RELAP5-3D, both the physics and numerical methods of the code have been continuously improved. In the area of computational methods and mathematical techniques, many upgrades and improvements have been made decrease code run time and increase solution accuracy. These include vectorization, parallelization, use of improved equation solvers for thermal hydraulics and neutron kinetics, and incorporation of improved library utilities. In the area of applied nuclear engineering, expanded capabilities include boron and level tracking models, radiation/conduction enclosure model, feedwater heater and compressor components, fluids and corresponding correlations for modeling Generation IV reactor designs, and coupling to computational fluid dynamics solvers. Ongoing and proposed future developments include improvements to the two-phase pump model, conversion to FORTRAN 90, and coupling to more computer programs. This paper summarizes the general improvements made to RELAP5-3D, with an emphasis on streamlining the code infrastructure for improved maintenance and development. With all these past, present and planned developments, it is necessary to modify the code infrastructure to incorporate modifications in a consistent and maintainable manner. Modifying a complex code such as RELAP5-3D to incorporate new models, upgrade numerics, and optimize existing code becomes more difficult as the code grows larger. The difficulty of this as well as the chance of introducing errors is significantly reduced when the code is structured. To streamline the code into a structured program, a commercial restructuring tool, FOR_STRUCT, was applied to the RELAP5-3D source files. The methodology employed follows Dijkstra's structured programming paradigm, which is based on splitting programs into sub-sections, each with single points of entry and exit and in which control is passed downward through the structure with no unconditional branches to higher levels. GO TO commands are typically avoided, since they alter the flow and control of a program's execution by allowing a jump from one place in the routine to another. The restructuring of RELAP5-3D subroutines is complicated by several issues. The first is use of code other than standard FORTRAN77. The second is restructuring limitations of FOR_STRUCT. The third is existence of pre-compiler directives and the complication of nested directives. Techniques were developed to overcome all these difficulties and more and these are reported. By implementing these developments, all subroutines of RELAP were restructured. Measures of code improvement relative to maintenance and development are presented.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 74
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This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.
Author: Ferry Roelofs
Publisher: Woodhead Publishing
ISBN: 0081019815
Category : Science
Languages : en
Pages : 464
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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
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Category : Nuclear power plants
Languages : en
Pages : 24
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Author: Richard G. Warren
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Category : Geochemical prospecting
Languages : en
Pages : 110
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Author: International Atomic Energy Agency
Publisher:
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Category : Business & Economics
Languages : en
Pages : 720
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Author: International Atomic Energy Agency
Publisher:
ISBN: 9789201025180
Category : Technology & Engineering
Languages : en
Pages : 0
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This Safety Guide provides recommendations and guidance on a general framework for performing prospective radiological impact assessments for facilities and activities, to estimate and control the radiological effects on the public and on the environment. This radiological environmental impact assessment is intended for planned exposure situations as part of the authorization process and, when applicable, as part of a governmental decision making process for facilities and activities. The situations covered in the assessment include both exposures expected to occur in normal operation as well as potential exposures. The assessment of the radiological impacts includes consideration of the risk of radiation effects for humans and for populations of non-human biota. Guidance is provided on the assumptions and input data to be used, the necessary models for environmental transfer and radiation dose assessment and the definition and use of criteria for informing decisions.
Author: Evgeny Adamov
Publisher: Academic Press
ISBN: 0323993095
Category : Business & Economics
Languages : en
Pages : 468
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Book Description
Closed Nuclear Fuel Cycle with Fast Reactors: Handbook of Russian Nuclear Power provides unique insights into research and practical activities from leading Russian experts. It presents readers with unprecedented insight and essential knowledge surrounding nuclear fast reactor technologies, as well as novel methods to close the nuclear fuel cycle to achieve cleaner, more environmentally friendly, and more efficient nuclear power. Using the Proryv Project as a framework, the book's contributors provide detailed descriptions of technologies in development in Russia, allowing readers from around the globe to gain a thorough understanding which they can then apply to their own research and practice. Nuclear engineers and technologists of fast reactors, advanced reactors and fuel cycles will use this book as a guide to inform new technology development. They will be able to use the experiences from the Proryv Project to drive fast reactor development with closed fuel cycles for the future. Provides a presentation of new nuclear reactor and fuel cycle technologies within the unique framework of Russia’s Proryv Project Presents novel technologies to close the nuclear fuel cycle to promote cleaner and more environmentally protective nuclear power Includes thorough coverage on the topic, including core design, coolants, fuels, accident protection and waste management technologies
Author: Neil E. Todreas
Publisher:
ISBN: 9781351030472
Category : Science
Languages : en
Pages : 926
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Book Description
Nuclear Systems, Volume I: Thermal Hydraulic Fundamentals, Third Edition, provides an in-depth introduction to nuclear power, focusing on thermal hydraulic design and analysis of the nuclear core and other key nuclear plant components. The authors stress the integration of fluid flow and heat transfer as applied to all power reactor types and energy source distribution. They cover nuclear reactor concepts and systems, including GEN III+, GEN IV, and SMR reactors and new power cycles. The text includes new chapter examples and problems using concept parameters, full-color text and art, computer programs, figure slides, and a solutions manual. FEATURES Rigorous coverage of nuclear power generation fundamentals Description and analysis of the latest nuclear power plant designs and technologies Extensive examples in each chapter to illustrate the analysis methods which have been presented New full-color art and text features to enhance the presentation of topics Integration of fluid flow and heat transfer as applied to single- and two-phase coolants Readers will develop the knowledge and design skills needed to improve the next generation of nuclear reactors.
