Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Modeling of deposition and erosion of CRUD on fuel surfaces under sub-cooled nucleate boiling in PWR

  • Seungjin Seo (Department of Nuclear Engineering, Seoul National University) ;
  • Nakkyu Chae (Department of Nuclear Engineering, Seoul National University) ;
  • Samuel Park (Department of Nuclear Engineering, Seoul National University) ;
  • Richard I. Foster (Nuclear Research Institute for Future Technology and Policy, Seoul National University) ;
  • Sungyeol Choi (Department of Nuclear Engineering, Seoul National University)
  • Received : 2022.11.30
  • Accepted : 2023.04.04
  • Published : 2023.07.25
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Abstract

Simulating the Corrosion-Related Unidentified Deposit (CRUD) on the surface of fuel assemblies is necessary to predict the axial offset anomaly and the localized corrosion induced by the CRUD during the operation of nuclear power plants. A new CRUD model was developed to predict the formation of the CRUD deposits, considering the deposition and erosion mechanisms. The heat transfer and capillary flow within the CRUD were also considered to evaluate the boiling amount within the CRUD layer. This model predicted a CRUD deposit thickness of 44 ㎛ during a one-cycle operation of the Seabrook nuclear power plant. The CRUD deposition tended to accelerate and decelerate during the simulation, by being related to boiling mechanism on the deposits surface. Additionally, during a three-cycle operation corresponding to the refueling period, the CRUD deposition was saturated at a thickness of 80 ㎛, which was in good agreement with the suggested thickness for CRUD buildupin pressurized water reactors. Surface boiling on the thin CRUD deposits enhanced the acceleration of the deposition, even when the wick boiling properties were not favorable for CRUD deposition. To ensure the certainty of the simulation results, sensitivity analyses were conducted for the porosity, chimney density, and the constants employed in the proposed model of the CRUD.

Keywords

Acknowledgement

This work was supported by the Nuclear Safety Research Program through the Korea Foundation Of Nuclear Safety (KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission (NSSC) of the Republic of Korea (No. 2106022), the Korea Institute of Energy Technology Evaluation and Planning (KETEP) and the Ministry of Trade, Industry & Energy (MOTIE) of the Republic of Korea (No. 20224000000120), and the National Research Foundation of Korea (NRF) grant funded by the Korean goverment (Ministry of Science, ICT and Future Planning) (No. RS202200144515).

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