OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests (PSBT) - Scoping Calculations with the Advanced Subchannel Code CTF
The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark was established to encourage advancement in subchannel analysis of fluid flow in rod bundles, which has a significant impact on nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for this international benchmark activity. The benchmark is designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and DNB to the full-scale experimental data. It consists of seven exercises grouped in two phases and is designed to determine the capabilities of the selected codes to perform steady-state and transient void distribution calculations for both a single subchannel and a full bundle, as well as steady-state pressure drop, steady-state and transient DNB, and fluid temperature calculations for a full bundle. The results provided by each code are to be compared not only to the measured data but also to the results from other codes in an attempt to determine the strengths and weaknesses of the models utilized in each code. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety (JNES) organization including the participation and support of the US Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU is performing supporting calculations of the benchmark exercises using its in-house advanced thermal-hydraulic subchannel code CTF. CTF is a version of the well known and widely used COBRA-TF code whose models have been continuously improved and validated at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU over the last years. This paper discusses the problem specification for OECD/NRC PSBT benchmark as well as CTF applications to selected benchmark exercises. The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark was established to encourage advancement in subchannel analysis of fluid flow in rod bundles, which has a significant impact on nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for this international benchmark activity. The benchmark is designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and DNB to the full-scale experimental data. It consists of seven exercises grouped in two phases and is designed to determine the capabilities of the selected codes to perform steady-state and transient void distribution calculations for both a single subchannel and a full bundle, as well as steady-state pressure drop, steady-state and transient DNB, and fluid temperature calculations for a full bundle. The results provided by each code are to be compared not only to the measured data but also to the results from other codes in an attempt to determine the strengths and weaknesses of the models utilized in each code. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety (JNES) organization including the participation and support of the US Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU is performing supporting calculations of the benchmark exercises using its in-house advanced thermal-hydraulic subchannel code CTF. CTF is a version of the well known and widely used COBRA-TF code whose models have been continuously improved and validated at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU over the last years. This paper discusses the problem specification for OECD/NRC PSBT benchmark as well as CTF applications to selected benchmark exercises. The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark was established to encourage advancement in subchannel analysis of fluid flow in rod bundles, which has a significant impact on nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for this international benchmark activity. The benchmark is designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and DNB to the full-scale experimental data. It consists of seven exercises grouped in two phases and is designed to determine the capabilities of the selected codes to perform steady-state and transient void distribution calculations for both a single subchannel and a full bundle, as well as steady-state pressure drop, steady-state and transient DNB, and fluid temperature calculations for a full bundle. The results provided by each code are to be compared not only to the measured data but also to the results from other codes in an attempt to determine the strengths and weaknesses of the models utilized in each code. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety (JNES) organization including the participation and support of the US Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU is performing supporting calculations of the benchmark exercises using its in-house advanced thermal-hydraulic subchannel code CTF. CTF is a version of the well known and widely used COBRA-TF code whose models have been continuously improved and validated at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU over the last years. This paper discusses the problem specification for OECD/NRC PSBT benchmark as well as CTF applications to selected benchmark exercises. The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark was established to encourage advancement in subchannel analysis of fluid flow in rod bundles, which has a significant impact on nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for this international benchmark activity. The benchmark is designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and DNB to the full-scale experimental data. It consists of seven exercises grouped in two phases and is designed to determine the capabilities of the selected codes to perform steady-state and transient void distribution calculations for both a single subchannel and a full bundle, as well as steady-state pressure drop, steady-state and transient DNB, and fluid temperature calculations for a full bundle. The results provided by each code are to be compared not only to the measured data but also to the results from other codes in an attempt to determine the strengths and weaknesses of the models utilized in each code. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety (JNES) organization including the participation and support of the US Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU is performing supporting calculations of the benchmark exercises using its in-house advanced thermal-hydraulic subchannel code CTF. CTF is a version of the well known and widely used COBRA-TF code whose models have been continuously improved and validated at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU over the last years. This paper discusses the problem specification for OECD/NRC PSBT benchmark as well as CTF applications to selected benchmark exercises. The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark was established to encourage advancement in subchannel analysis of fluid flow in rod bundles, which has a significant impact on nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for this international benchmark activity. The benchmark is designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and DNB to the full-scale experimental data. It consists of seven exercises grouped in two phases and is designed to determine the capabilities of the selected codes to perform steady-state and transient void distribution calculations for both a single subchannel and a full bundle, as well as steady-state pressure drop, steady-state and transient DNB, and fluid temperature calculations for a full bundle. The results provided by each code are to be compared not only to the measured data but also to the results from other codes in an attempt to determine the strengths and weaknesses of the models utilized in each code. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety (JNES) organization including the participation and support of the US Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU is performing supporting calculations of the benchmark exercises using its in-house advanced thermal-hydraulic subchannel code CTF. CTF is a version of the well known and widely used COBRA-TF code whose models have been continuously improved and validated at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU over the last years. This paper discusses the problem specification for OECD/NRC PSBT benchmark as well as CTF applications to selected benchmark exercises. The international OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark was established to encourage advancement in subchannel analysis of fluid flow in rod bundles, which has a significant impact on nuclear reactor safety margin evaluations. The benchmark is based on one of the most valuable databases identified for the thermal-hydraulics modeling, which was developed by the Nuclear Power Engineering Corporation (NUPEC) in Japan. The database includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurized Water Reactor (PWR) fuel assembly. Part of this database is made available for this international benchmark activity. The benchmark is designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and DNB to the full-scale experimental data. It consists of seven exercises grouped in two phases and is designed to determine the capabilities of the selected codes to perform steady-state and transient void distribution calculations for both a single subchannel and a full bundle, as well as steady-state pressure drop, steady-state and transient DNB, and fluid temperature calculations for a full bundle. The results provided by each code are to be compared not only to the measured data but also to the results from other codes in an attempt to determine the strengths and weaknesses of the models utilized in each code. The PSBT benchmark team is organized based on the collaboration between the Pennsylvania State University (PSU) and the Japan Nuclear Energy Safety (JNES) organization including the participation and support of the US Nuclear Regulatory Commission (NRC) and the Nuclear Energy Agency (NEA), OECD. On behalf of the PSBT benchmark team, PSU is performing supporting calculations of the benchmark exercises using its in-house advanced thermal-hydraulic subchannel code CTF. CTF is a version of the well known and widely used COBRA-TF code whose models have been continuously improved and validated at the Reactor Dynamics and Fuel Management Group (RDFMG) at PSU over the last years. This paper discusses the problem specification for OECD/NRC PSBT benchmark as well as CTF applications to selected benchmark exercises.
PSBT Benchmark CTF OECD/NRC NUPEC
A. Rubin M. Avramova H. Utsuno
Nuclear Engineering Program The Pennsylvania State University University Park, PA 16802, USA Japan Nuclear Energy Safety Organization Kamiya-Cho MT Bldg., 4-3-20, Toranomon, Minato-ku
国际会议
上海
英文
2142-2151
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）