BWR small break LOCA counterpart tests at ROSA-III and FIST test facilities

Nuclear Engineering and Design - Tập 102 - Trang 151-163 - 1987
Y. Koizumi1, H. Nakamura1, K. Tasaka1, J.A. Findlay1, L.S. Lee1, W.A. Sutherland1
1Department of Reactor Safety Research, Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-11 Japan

Tài liệu tham khảo

Anode, 1981, ROSA-III system description for fuel assembly No. 4 Stephens, 1982, BWR full integral simulation program Tasaka, 1982, ROSA-III base test series for a large break loss-of-coolant accident in a boiling water reactor, Nucl. Technol., 57, 179, 10.13182/NT82-A26280 Tasaka, 1985, ROSA-III double-ended break test series for a loss-of-cooland accident in a boiling water reactor, Nucl. Technol., 68, 77, 10.13182/NT85-A33569 Hwang, 1983, BWR full integral simulation test (FIST) phase 1 test results Tasaka, 1984, ROSA-III and FIST BWR simulation tests: loss-of-coolant accident response comparisons, NEDC-30895 Tasaka, 1985, Comparisons of ROSA-III and FIST BWR loss of Cooland Accident Simulation Tests Suzuki, 1984, Recirculation pump suction line 2.8% break integral test at ROSA-III with HPCS failure Muramatsu, 1982, THYDE-B1/Modl: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors Wilson, 1962, The velocity of rising steam in a bubbling two-phase mixture, Trans. Am. Nucl. Soc., 5, 151 Cunningham, 1973, Experiments and void correlation for PWR small-break LOCA conditions, Trans. Am. Nucl. Soc., 17, 369 Moore, 1975, RELAP4 - a computer program for transient thermal hydraulic analysis T. Yonomoto et al., Heat transfer analysis at cladding surface during BWR LOCA experiment at ROSA-III, to be published in Nucl. Engrg. Des. Moody, 1965, Maximum flow rate of a single component, two-phase mixture, J. Heat Transfer, 87, 434