Abstract
The Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized the Organisation for Economic Co-operation and Development/US Nuclear Regulatory Commission (OECD/NRC) benchmark based on the Nuclear Power Engineering Corporation (NUPEC) pressurized water reactor (PWR) subchannel and bundle tests (PSBTs). The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency (NEA) of OECD and the Japan Nuclear Energy Safety Organization (JNES), Japan. The OECD/NRC PSBT benchmark was organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFDs) codes. The benchmark was designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and department from nucleate boiling (DNB), under steady-state and transient conditions, to full-scale experimental data. This paper provides an overview of the objectives of the benchmark along with a definition of the benchmark phases and exercises. The NUPEC PWR PSBT facility and the specific methods used in the void distribution measurements are discussed followed by a summary of comparative analyses of submitted final results for the exercises of the two benchmark phases.
Highlights
The need to refine the models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear engineering
One of the most valuable databases identified for the thermal-hydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan, which includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative pressurized water reactor (PWR) fuel assembly [1–3]
Part of this database is made available for an international benchmark activity entitled as the OECD/Nuclear Regulatory Commission (NRC) NUPEC PWR subchannel and bundle tests (PSBT) benchmark [4]
Summary
The need to refine the models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear engineering. One of the most valuable databases identified for the thermal-hydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan, which includes subchannel void fraction and departure from nucleate boiling (DNB) measurements in a representative pressurized water reactor (PWR) fuel assembly [1–3]. Void fraction measurements and departure from nucleate boiling (DNB) tests were performed at NUPEC under steady-state and transient pressurized water reactor (PWR) conditions Part of this database is made available for an international benchmark activity entitled as the OECD/NRC NUPEC PWR subchannel and bundle tests (PSBT) benchmark [4]. The exercises in phase I of the benchmark are designed to test the codes’ ability to predict void distribution in a single subchannel and a bundle under both steady-state and transient conditions as well as to calculate the pressure drop across a bundle. The exercises in Phase II of the benchmark are designed to test the codes’ ability to predict DNB in a bundle assembly under both steady-state and transient conditions, as well as to predict fluid temperatures under these conditions
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