Department of Nuclear Engineering Purdue University 400 Central Drive, West Lafayette, IN 47907-2017, USA;
Department of Nuclear Engineering Purdue University 400 Central Drive, West Lafayette, IN 47907-2017, USA;
Department of Nuclear Engineering Purdue University 400 Central Drive, West Lafayette, IN 47907-2017, USA;
Department of Nuclear Engineering Purdue University 400 Central Drive, West Lafayette, IN 47907-2017, USA;
Department of Nuclear Engineering Purdue University 400 Central Drive, West Lafayette, IN 47907-2017, USA;
Department of Nuclear Engineering Purdue University 400 Central Drive, West Lafayette, IN 47907-2017, USA;
Nuclear Fuel Industries Ltd.3135-41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki 311-1196, Japan;
Institute of Nuclear Safety System, Inc. 64, Sata, Mihama-cho, Mikata-gun, Fukui 919-1205, Japan;
Institute of Nuclear Safety System, Inc. 64, Sata, Mihama-cho, Mikata-gun, Fukui 919-1205, Japan;
rod bundle; two-phase flow; void fraction; drift-flux model; boiling water reactor;
机译:棒束几何形状的流通量相关性
机译:高压部件下杆束几何中的空隙分数.:漂移 - 助焊模型评估和发展
机译:棒束几何子通道中的漂移通量模型
机译:杆束几何的漂移通量相关性
机译:一维两相流漂移-通量闭合关系的可伸缩性研究,可用于RELAP5棒束和新的,比例良好的低液流量棒束数据。
机译:评估超临界水反应堆拟议棒束几何形状子通道内的传热相关性
机译:水平棒束在液态钠中的自然对流传热。第2部分:基于理论结果的水平棒束的相关性
机译:棒束几何中的稳态薄膜沸腾数据和非平衡相关评估