Development of a thermal-hydraulic analysis code for the Pebble Bed Water-cooled Reactor

作     者：Cai, Xiaoyu Qiu, Suizheng Tian, Wenxi Su, Guanghui 

作者机构：Xi An Jiao Tong Univ Sch Nucl Sci & Technol Xian 710049 Shaanxi Peoples R China Xi An Jiao Tong Univ State Key Lab Multiphase Flow Power Engn Xian 710049 Shaanxi Peoples R China 

出 版 物：《NUCLEAR ENGINEERING AND DESIGN》 (核工程与设计)

年 卷 期：2011年第241卷第12期

页      面：4978-4988页

核心收录：

学科分类：08[工学] 0827[工学-核科学与技术] 

主　　题：assemblies WATER COOLED REACTORS Safe handling core SAFETY MARGINS coolants Reactor cores flow field major loop REACTOR OPERATION heat Pebble bed Codes CRITICAL HEAT FLUX 

摘      要：The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly. Steady-state thermal-hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper. (C) 2011 Elsevier B.V. All rights reserved.