Nuclear Engineering and Technology

  • 제48권5호
  • /
  • Pages.1071-1082
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  • 2016
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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A Preliminary Safety Analysis for the Prototype Gen IV Sodium-Cooled Fast Reactor

  • 투고 : 2016.08.16
  • 심사 : 2016.08.17
  • 발행 : 2016.10.25
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초록

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

키워드

참고문헌

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  2. H.Y. Jeong, K.S. Ha, W.P. Chang, Y.M. Kwon, K.L. Lee, Thermal-Hydraulic Model in MARS-LMR, Rep. No. KAERI/TR-4297, Korea Atomic Energy Research Institute, Daejeon, Republic of Korea, 2011.
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  4. K.S. Ha, H.-Y. Jeong, W.-P. Chang, Y.-M. Kwon, C. Cho, Y.-B. Lee, Development of the MATRA-LMR-FB for flow blockage analysis in a LMR, Nucl. Eng. Tech 41 (2009) 797-806. https://doi.org/10.5516/NET.2009.41.6.797
  5. Toshiba Corporation, 4S Safety Analysis, Toshiba Technical Report, 2009. AFT-2009-000155 Rev000(0).
  6. L. Soffer, S.B. Burson, C.M. Ferrell, R.Y. Lee, J.N. Ridgely, Accident Source Terms for Light-Water Nuclear Power Plants (Rep. No. NUREG-1465), U.S. Nuclear Regulatory Commission, Washington, D.C, 1995.
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피인용 문헌

  1. Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor vol.322, pp.None, 2016, https://doi.org/10.1016/j.nucengdes.2017.06.049
  2. Experimental studies on eutectic formation between metallic fuel and HT-9M cladding in a single-pin core structure of a sodium-cooled fast reactor vol.505, pp.None, 2016, https://doi.org/10.1016/j.jnucmat.2018.04.003
  3. Performance evaluation of the Floating Absorber for Safety at Transient (FAST) in the innovative Sodium-cooled Fast Reactor (iSFR) under a single control rod withdrawal accident vol.52, pp.6, 2016, https://doi.org/10.1016/j.net.2019.11.011
  4. Experimental studies on metallic fuel relocation in a pin bundle core structure of a sodium-cooled fast reactor vol.365, pp.None, 2016, https://doi.org/10.1016/j.nucengdes.2020.110719
  5. Experimental validation of simulating natural circulation of liquid metal using water vol.52, pp.9, 2016, https://doi.org/10.1016/j.net.2020.03.005