T. Del Moro , P. Cioli Puviani , B. Gonfiotti , I. Di Piazza , D. Martelli , C. Ciurluini , F. Giannetti , R. Zanino , M. Tarantino
{"title":"通过新型 CFX-RELAP5 代码耦合分析在 NACIE-UP 设施进行的实验测试","authors":"T. Del Moro , P. Cioli Puviani , B. Gonfiotti , I. Di Piazza , D. Martelli , C. Ciurluini , F. Giannetti , R. Zanino , M. Tarantino","doi":"10.1016/j.nucengdes.2024.113676","DOIUrl":null,"url":null,"abstract":"<div><div>The design and safety assessment of Lead-cooled Fast Reactors (LFRs), being one of the Generation IV technologies, must be supported by extensive experimental campaigns. Such activities are necessary to completely understand the physical phenomena involved in such reactors, as well as to properly develop new numerical tools or validate the pre-existent ones. From the experimental point of view, ENEA Research Center of Brasimone is one of the most active institutions, thanks to its experimental platforms and know-how maturated since the early 2000s. From the numerical point of view, Computational Fluid Dynamics (CFD) codes are the most suitable ones to analyze some phenomena expected in a Heavy Liquid Metal (HLM)-cooled reactor, such as the complex 3D phenomena occurring within the pools or the core fuel assemblies. In addition, the fluid thermal conduction, usually neglected in a System Thermal-Hydraulic (STH) code, can assume a significant importance in some transient scenarios, e.g., loss of flow accidents with transition from forced to natural circulation. However, the safety analysis of the LFRs should still rely on the use of STH codes because of their lower computational cost compared to the CFD codes, also considering the high number of transient evolutions to be analyzed for the purpose of the reactor licensing. At ENEA Brasimone, a novel coupling approach has been developed to couple the CFD code Ansys CFX with the STH code RELAP5/Mod3.3. The coupled tool aims at exploiting the advantages of the two families of codes. It adopts a multi-scale approach to simulate in detail some circuit components while performing system-level analysis, so as to keep an acceptable computational time. The coupling technique is based on ad-hoc user routines written in FORTRAN and implemented in Ansys CFX, which acts as the master code. The user routines take care of time step management, data exchange, RELAP5 execution, and error checking. The goal of this paper is to assess the simulation capabilities of the coupled tool by reproducing a forced-to-natural-circulation transition test, carried out at the NACIE-UP facility, with LBE as working fluid. The work has been realized in the framework of the IAEA Coordinate Research Project-I31038, named “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop”.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113676"},"PeriodicalIF":1.9000,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":"{\"title\":\"Analysis of the experimental tests performed at NACIE-UP facility through a novel CFX-RELAP5 codes coupling\",\"authors\":\"T. Del Moro , P. Cioli Puviani , B. Gonfiotti , I. Di Piazza , D. Martelli , C. Ciurluini , F. Giannetti , R. Zanino , M. Tarantino\",\"doi\":\"10.1016/j.nucengdes.2024.113676\",\"DOIUrl\":null,\"url\":null,\"abstract\":\"<div><div>The design and safety assessment of Lead-cooled Fast Reactors (LFRs), being one of the Generation IV technologies, must be supported by extensive experimental campaigns. Such activities are necessary to completely understand the physical phenomena involved in such reactors, as well as to properly develop new numerical tools or validate the pre-existent ones. From the experimental point of view, ENEA Research Center of Brasimone is one of the most active institutions, thanks to its experimental platforms and know-how maturated since the early 2000s. From the numerical point of view, Computational Fluid Dynamics (CFD) codes are the most suitable ones to analyze some phenomena expected in a Heavy Liquid Metal (HLM)-cooled reactor, such as the complex 3D phenomena occurring within the pools or the core fuel assemblies. In addition, the fluid thermal conduction, usually neglected in a System Thermal-Hydraulic (STH) code, can assume a significant importance in some transient scenarios, e.g., loss of flow accidents with transition from forced to natural circulation. However, the safety analysis of the LFRs should still rely on the use of STH codes because of their lower computational cost compared to the CFD codes, also considering the high number of transient evolutions to be analyzed for the purpose of the reactor licensing. At ENEA Brasimone, a novel coupling approach has been developed to couple the CFD code Ansys CFX with the STH code RELAP5/Mod3.3. The coupled tool aims at exploiting the advantages of the two families of codes. It adopts a multi-scale approach to simulate in detail some circuit components while performing system-level analysis, so as to keep an acceptable computational time. The coupling technique is based on ad-hoc user routines written in FORTRAN and implemented in Ansys CFX, which acts as the master code. The user routines take care of time step management, data exchange, RELAP5 execution, and error checking. The goal of this paper is to assess the simulation capabilities of the coupled tool by reproducing a forced-to-natural-circulation transition test, carried out at the NACIE-UP facility, with LBE as working fluid. 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Analysis of the experimental tests performed at NACIE-UP facility through a novel CFX-RELAP5 codes coupling
The design and safety assessment of Lead-cooled Fast Reactors (LFRs), being one of the Generation IV technologies, must be supported by extensive experimental campaigns. Such activities are necessary to completely understand the physical phenomena involved in such reactors, as well as to properly develop new numerical tools or validate the pre-existent ones. From the experimental point of view, ENEA Research Center of Brasimone is one of the most active institutions, thanks to its experimental platforms and know-how maturated since the early 2000s. From the numerical point of view, Computational Fluid Dynamics (CFD) codes are the most suitable ones to analyze some phenomena expected in a Heavy Liquid Metal (HLM)-cooled reactor, such as the complex 3D phenomena occurring within the pools or the core fuel assemblies. In addition, the fluid thermal conduction, usually neglected in a System Thermal-Hydraulic (STH) code, can assume a significant importance in some transient scenarios, e.g., loss of flow accidents with transition from forced to natural circulation. However, the safety analysis of the LFRs should still rely on the use of STH codes because of their lower computational cost compared to the CFD codes, also considering the high number of transient evolutions to be analyzed for the purpose of the reactor licensing. At ENEA Brasimone, a novel coupling approach has been developed to couple the CFD code Ansys CFX with the STH code RELAP5/Mod3.3. The coupled tool aims at exploiting the advantages of the two families of codes. It adopts a multi-scale approach to simulate in detail some circuit components while performing system-level analysis, so as to keep an acceptable computational time. The coupling technique is based on ad-hoc user routines written in FORTRAN and implemented in Ansys CFX, which acts as the master code. The user routines take care of time step management, data exchange, RELAP5 execution, and error checking. The goal of this paper is to assess the simulation capabilities of the coupled tool by reproducing a forced-to-natural-circulation transition test, carried out at the NACIE-UP facility, with LBE as working fluid. The work has been realized in the framework of the IAEA Coordinate Research Project-I31038, named “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop”.
期刊介绍:
Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology.
Fundamentals of Reactor Design include:
• Thermal-Hydraulics and Core Physics
• Safety Analysis, Risk Assessment (PSA)
• Structural and Mechanical Engineering
• Materials Science
• Fuel Behavior and Design
• Structural Plant Design
• Engineering of Reactor Components
• Experiments
Aspects beyond fundamentals of Reactor Design covered:
• Accident Mitigation Measures
• Reactor Control Systems
• Licensing Issues
• Safeguard Engineering
• Economy of Plants
• Reprocessing / Waste Disposal
• Applications of Nuclear Energy
• Maintenance
• Decommissioning
Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.