Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Verification of neutronics and thermal-hydraulic coupled system with pin-by-pin calculation for PWR core

  • Zhigang Li (Nuclear Power Institute of China) ;
  • Junjie Pan (Nuclear Power Institute of China) ;
  • Bangyang Xia (Science and Technology on Reactor System Design Technology Laboratory) ;
  • Shenglong Qiang (Nuclear Power Institute of China) ;
  • Wei Lu (Nuclear Power Institute of China) ;
  • Qing Li (Nuclear Power Institute of China)
  • Received : 2022.09.07
  • Accepted : 2023.04.12
  • Published : 2023.09.25
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Abstract

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO2. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

Keywords

References

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