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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.
Sunil Pak, Mun-Seong Cheon, Hyeon Gon Lee, Michael R. Kalish, C. S. Pitcher, Christopher I. Walker
Fusion Science and Technology | Volume 56 | Number 1 | July 2009 | Pages 129-133
Plasma Engineering and Diagnostics | Eighteenth Topical Meeting on the Technology of Fusion Energy (Part 1) | doi.org/10.13182/FST09-A8889
Articles are hosted by Taylor and Francis Online.
A preliminary thermo-hydraulic analysis was performed on the ITER diagnostic upper port plug. Relevant thermal and hydraulic parameters, such as coolant pressure drop, maximum structure temperature and bake-out time, were calculated for normal operation and baking. The upper port plug considered is based on the preliminary generic structure design of Princeton Plasma Physics Laboratory and the Blanket Shield Module (BSM) developed in Europe. The diagnostic shield modules are modeled so that the Korean diagnostic procurement package, which includes Vacuum Ultra-Violet (VUV) spectrometer and neutron activation system, can be integrated. The analysis provides design inputs to optimize flow in the cooling channels of the plug. The conjugated heat transfer analysis for the port plug confirms that it is important to secure accurate nuclear heat and accurate electro-magnetic (EM) force for the design of the joining flange between the BSM and the main body. Thermal analysis shows that it will take ten hours for the port plug to reach the bake-out temperature (240°C), if the window plate is heated additionally from the rear side.