Numerical assessment of NaCl-KCl-UCl3 spreading and solidification behavior in a salt spill accident

  • Chun, Ye Hwan
  • Park, Jae Hyung
  • Cho, Yun Sik
  • An, Sang Mo
  • Kim, Sung Il
  • ... Kim, Sung Joong
  • 외 1명
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초록

Molten salt reactors (MSRs) are emerging as next-generation nuclear systems due to their enhanced safety characteristics. Accurate prediction of molten salt spreading and solidification is essential for regulatory approval, as these phenomena govern potential radioactive release. This study develops a CFD framework to simulate the spreading of NaCl–KCl–UCl3 salt using the Volume of Fluid and enthalpy–porosity methods to capture multiphase flow and phase change. A systematic sensitivity analysis is conducted to identify dominant physical parameters. Results show that mushy zone constant is the most influential factor, with variations altering the spreading radius by up to 53% due to changes in flow resistance during solidification. Turbulence modeling introduces up to a 13% variation in heat removal, while radiative heat transfer has a negligible effect. Additional parametric studies indicate that increasing mass flow rate proportionally enlarges both spreading area and total heat dissipation, whereas initial spill temperature has minimal influence on spreading behavior. These findings highlight the critical importance of reducing uncertainty in thermos-physical properties, particularly the mushy zone parameter, for reliable modeling. The proposed framework provides a systematic basis for model selection and supports the verification and validation of salt spill accident behavior as part of regulatory processes for MSR deployment.

키워드

Computational fluid dynamicsMolten salt fuel spillMolten salt reactorPhase-changeSolidification-meltingVISCOSITY
제목
Numerical assessment of NaCl-KCl-UCl3 spreading and solidification behavior in a salt spill accident
저자
Chun, Ye HwanPark, Jae HyungCho, Yun SikAn, Sang MoKim, Sung IlSong, JinHoKim, Sung Joong
DOI
10.1016/j.net.2026.104358
발행일
2026-08
유형
Article
저널명
Nuclear Engineering and Technology
58
8
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1 ~ 10