Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Residual stress distribution analysis in a J-groove dissimilar metal welded component of a reactor vessel bottom head using simulation and experiment

  • Dong-Hyun Ahn (Materials Safety Technology Research Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Jong Yeon Lee (Materials Safety Technology Research Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Min-Jae Choi (Materials Safety Technology Research Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Jong Min Kim (Materials Safety Technology Research Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Sung-Woo Kim (Materials Safety Technology Research Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Wanchuck Woo (Neutron Science Division, Korea Atomic Energy Research Institute (KAERI))
  • Received : 2023.07.02
  • Accepted : 2023.10.20
  • Published : 2024.02.25
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Abstract

To simulate the verification process using materials from a decommissioned reactor, a mock-up of the bottom-mounted instrument nozzle in the Kori 1 reactor, where the nozzle was attached to a plate by J-groove dissimilar metal welding, was fabricated. The mock-up distortion was quantified by measuring the plate surface displacement after welding. The residual stresses formed on the support plate surface and the inner surface of the nozzle were then analyzed using the hole-drilling method, contour method, and neutron diffraction. Welding simulations were performed using a 3D finite element method to validate the measured results. The measured and computed stress distributions on the support plate exhibited reasonable agreement. Conversely, the stresses on the inside of the nozzle were found to have an indisputable difference in the contour method and neutron diffraction measurements, which demonstrated strong tensile and compressive hoop stresses, respectively. The possible origins of such differences were investigated and we have provided some suggestions for a precise evaluation in the simulation. This study is expected to be useful in future research on decommissioned reactors.

Keywords

Acknowledgement

This work was supported by the Korea Atomic Energy Research Institute R&D Program (Contract No. 524480-23) and the National Research Foundation of Korea (NRF) grant funded by the Korean Government (2021M2E4A1037979).

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