Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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High fidelity transient solver in STREAM based on multigroup coarse-mesh finite difference method

  • Anisur Rahman (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology) ;
  • Hyun Chul Lee (School of Mechanical Engineering, Pusan National University) ;
  • Deokjung Lee (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology)
  • Received : 2022.09.29
  • Accepted : 2023.05.14
  • Published : 2023.09.25
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Abstract

This study incorporates a high-fidelity transient analysis solver based on multigroup CMFD in the MOC code STREAM. Transport modeling with heterogeneous geometries of the reactor core increases computational cost in terms of memory and time, whereas the multigroup CMFD reduces the computational cost. The reactor condition does not change at every time step, which is a vital point for the utilization of CMFD. CMFD correction factors are updated from the transport solution whenever the reactor core condition changes, and the simulation continues until the end. The transport solution is adjusted once CMFD achieves the solution. The flux-weighted method is used for rod decusping to update the partially inserted control rod cell material, which maintains the solution's stability. A smaller time-step size is needed to obtain an accurate solution, which increases the computational cost. The adaptive step-size control algorithm is robust for controlling the time step size. This algorithm is based on local errors and has the potential capability to accept or reject the solution. Several numerical problems are selected to analyze the performance and numerical accuracy of parallel computing, rod decusping, and adaptive time step control. Lastly, a typical pressurized LWR was chosen to study the rod-ejection accident.

Keywords

Acknowledgement

This work was partially supported by a National Research Foundation of Korea (NRF) grant funded by the Korean government (MSIT) (No. NRF-2019M2D2A1A03058371) and the project (L20S089000) by Korea Hydro and Nuclear Power Co. Ltd. This work was partially supported by Korea Institute of Energy Technology Evaluation and Planning (KETEP) grant funded by the Korea government (MOTIE) [RS-2023-00241302]. The authors would like to express their sincere gratitude to Brendan Kochunas, Assistant Professor, Department of Nuclear Engineering and Radiological Sciences, University of Michigan, and Qicang Shen, Post-doctoral Researcher, the University of Michigan, for providing the MPACT results.

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