题目：Inverted Annual Film Boiling of Water in a Vertical Tubular Test Section
时间：2018年7月2日 9：30-11：00
地点：德赢vwin055 F310会议室
邀请人：尹俊连副研究员（核科学与工程学院）
Biography
Xiaodong Sun is Professor of Nuclear Engineering and Radiological Sciences at the University of Michigan. Prior to joining UM, Dr. Sun was Professor of Nuclear and Mechanical Engineering at The Ohio State University. He received his B.S. degree in Thermal Engineering and MSNE both from Shanghai Jiao Tong University; and MSME and Ph.D. in Nuclear Engineering from Purdue University. Dr. Sun continued at Purdue University as a postdoctoral researcher before he joined OSU in 2004. Dr. Sun’s research interest areas include reactor thermal hydraulics and safety; two-phase flow experimentation and modeling; and thermal hydraulics and heat exchangers for advanced nuclear reactors. Dr. Sun has co-authored over 210 peer reviewed journal articles and conference papers. He was chair of the American Nuclear Society’s Thermal Hydraulics Division (2012-2013) and chair of the Division’s Program Committee (2013-2016). Dr. Sun served/serves as the Technical Program Chair (TPC) for the ANS 2012 Winter Meeting, and co-TPC for NUTHOS-10 (2014), NURETH-16 (2015), NUTHOS-11 (2016), and NUTHOS-12 (2018).
Abstract
Post critical heat flux (post-CHF) heat transfer may occur during loss of coolant accidents (LOCAs) in water-cooled nuclear reactors. The post-CHF flow regimes include inverted annular film boiling (IAFB), inverted slug film boiling (ISFB), and dispersed flow film boing (DFFB). To better understand the characteristics of the hydrodynamics and heat transfer in the post-CHF flow regimes at high-pressure and high-flow conditions, a Post-CHF Heat Transfer (PCHT) test facility with a tubular test section using water as the working fluid was designed to perform quasi-steady-state film boiling experiments by using the directly-heated hot patch technique.
This presentation will be divided into two parts. The first part will offer a brief introduction to the research activities at the Thermal Hydraulics Laboratory (THL) in the Department of Nuclear Engineering and Radiological Sciences at the University of Michigan. The second part of the talk will focus on the research carried out on post-CHF heat transfer at THL. The design of the PCHT test facility will be introduced, including COMSOL Multiphysics analyses that helped inform the test section design, including the hot patch design and required heating power. In addition, a computer code was developed for the IAFB flow regime based on a one-dimensional two-fluid model to calculate the wall temperature, liquid core temperature, void fraction, and pressure drop, which provided the basis for identifying the IAFB test matrix. Finally, updates on the PCHT test facility construction and shakedown tests will be given.
