Nuclear Engineering and Technology

  • 제54권6호
  • /
  • Pages.2188-2197
  • /
  • 2022
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Thermal stress intensity factor solutions for reactor pressure vessel nozzles

  • Jeong, Si-Hwa (School of Mechanical Engineering, Sungkyunkwan University) ;
  • Chung, Kyung-Seok (School of Mechanical Engineering, Sungkyunkwan University) ;
  • Ma, Wan-Jun (School of Mechanical Engineering, Sungkyunkwan University) ;
  • Yang, Jun-Seog (Central Research Institute, Korea Hydro & Nuclear Power Co, Ltd) ;
  • Choi, Jae-Boong (School of Mechanical Engineering, Sungkyunkwan University) ;
  • Kim, Moon Ki (School of Mechanical Engineering, Sungkyunkwan University)
  • 투고 : 2021.08.22
  • 심사 : 2022.01.03
  • 발행 : 2022.06.25
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초록

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

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과제정보

This research was supported by Korea Hydro & Nuclear Power Co., Ltd. (G18IO11).

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