Critical Heat Flux Experiments for In-Vessel Retention External Reactor Vessel Cooling Strategy using 2-D Slice Test Section
The critical heat flux (CHF) on the reactor vessel external wall was measured using the small scale two-dimensional slice test section. The radius of the curvature and the channel area of the test section were 0.15 m and 3 cm×3 cm, respectively. The channel area and the heater width as well as the gap size are relatively very small comparing with the ULPU and KAIST-CHF experiments. The objectives are to measure the main characteristics of two-dimensional, two phase flow in order to evaluate the CHF data for a wide range of thermo-hydraulic (material of test section, SUS304; inlet subcooling, 2～10K; mass flux, 50～300 kg/m2s) and geometric (inclination angle 90°) parameters and to compare the CHF data of this study with the CHF data of large scale two-dimensional experiments such as KAIST-CHF and ULPU experiments. The materials of the test section were Type 304 stainless steel which is same with the KAIST-CHF. The inlet and outlet part of the test section simulated the APR1400 design. In aspect of exit quality, the CHF data compared with the SULTAN correlation.
IVR-ERVC CHF small scale 2-D Slice
Hae Min Park Taeil Kim Yong Hoon Jeong Sun Heo
Nuclear and Quantum Eng. Korea Advanced Institute of Science and Technology 335 Gwahak-ro, Yuseong-g Nuclear Engineering and Technology Institute, Korea Hydro & Nuclear Power Co.25-1, Jang-dong, Yuseon
国际会议
上海
英文
2036-2044
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）