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Spent fuel recycling and conditioning topic of U.S.-Japan meeting
Officials with the Department of Energy’s Office of Environmental Management discussed spent nuclear fuel recycling and conditioning with counterparts from Japan during the 13th U.S.-Japan Technical Meeting of the Civil Nuclear Energy Research and Development Working Group, held recently in Santa Fe, N.M.
Jin-Seok Hwang, Jong-Won Kim, Heon-Uk Nam, Goon-Cherl Park
Nuclear Technology | Volume 176 | Number 2 | November 2011 | Pages 260-273
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT11-A13300
Articles are hosted by Taylor and Francis Online.
A major safety factor in marine reactor design, critical heat flux (CHF), is assessed using the MARS system analysis code under heaving conditions. As gravity acceleration changes, the CHF is affected by the thermal hydraulics in the reactor through inlet flow fluctuations. Performing the analysis with the MARS code, which uses the properties of water for the working fluid, requires applying the CHF experimental data using fluid-to-fluid (FTF) scaling because most CHF experiments are conducted with Freon (R-113) as the working fluid. The FTF scaling methods suggested by Ahmad, Katto, and Coffield are adopted and compared. Otsuji et al.'s experiment, which was conducted using mass flow rate oscillation, is applied to evaluate the capability of MARS for heaving conditions. According to the calculations the FTF methods of Ahmad, Katto, and Coffield show good agreement (within an error of ±10.73% for Otsuji et al.'s experiment) for inlet flow rate oscillation corresponding to gravity acceleration in a vertical direction. In addition, variation of the acceleration affects the flow conditions, such as the mass flow rate and the void fraction. Thus, MARS has a noteworthy ability to predict the CHF for heaving conditions by simulating inlet flow rate oscillation.