한국가시화정보학회지 (Journal of the Korean Society of Visualization)

  • 제9권4호
  • /
  • Pages.54-62
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  • 2011
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  • 1598-8430(pISSN)
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  • 2093-808X(eISSN)
한국가시화정보학회 (The Korean Society of Visualization)
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소듐냉각고속로 KALIMER-600 축소 물모의 열유동 가시화 실험장치 구축 및 거시 유동장 특성 측정

Water-Simulant Facility Installation for the Sodium-Cooled Fast Reactor KALIMER-600 and Global Flow Measurement

  • 차재은 (한국원자력연구원 고속로기술개발부) ;
  • 김성오 (한국원자력연구원 고속로기술개발부)
  • 투고 : 2011.12.12
  • 심사 : 2011.12.28
  • 발행 : 2011.12.31
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초록

KAERI has developed a KALIMER-600 which is a pool-type sodium-cooled fast reactor with a 600MWe electric generation capacity. For a SFR development, one of the main topics is an enhancement of the reactor system safety. Therefore, we have a long-term plan to design the large sodium experimental facility to evaluate the reactor safety and component performance. In order to extrapolate a thermal hydraulic phenomena in a large sodium reactor, the thermal hydraulics phenomena is under investigation in a 1/$10^{th}$ water-simulant facility for the KALIMER-600. In this paper, we shortly described the experimental facility setup and the measurement of the isothermal global flow behavior. For the flow field measurement, the PIV method was used in a transparent Plexiglas reactor vessel model at around $20^{\circ}C$ water condition.

키워드

참고문헌

  1. D.H. Han, et al., 2007, KALIMER-600 Conceptual Design Report, KAERI/TR-3381/2007, KAERI Technical Report
  2. Hoffmann, H., Marten, K., Weinberg, D. and Ieda, Y., 1989, "Investigations on Natural Circulation in Reactor Models and Shut Down Heat Removal Systems for LMFBR's," Proc. 5th Nuclear Thermal Hydraulics, San Francisco, California
  3. Hoffmann, H., Marten, K., Weinberg, D. and Kamide, H., 1990, "Thermohydraulic Model Experiments and Calculations on the Transition from Forced to Natural Circulation for Pool-Type Fast Reactors," Transactions of the American Nuclear Society, Vol. 62, pp. 143-151
  4. N. Kimura, K. Hayashi, H. Kamide, M. Itoh, and T. Sekine, 2005, Experimental Study on Flow Optimization in Upper Plenum of Reactor Vessel for a Compact Sodium-Cooled Fast Reactor, Vol. 152(2), pp. 210-222 https://doi.org/10.13182/NT05-A3671