第36卷第2期 2016年4月
核科学与工程
Nuclear Science and Engineering
Vol.36No.2
Apr.2016
基于威斯康辛冷凝试验本体的结构改进及分析
杨林
（国家核电技术有限公司北京研发中心，北京100190）
摘要：先进压水堆（APWR)是第三代核电技术的代表堆型之一，它采用了非能动安全系统，提高了安全性能。非能动安全壳冷却系统(PCCS)主要利用蒸汽的冷凝来带走安全壳内的热量：本文主要介绍了威斯康辛大学进行的冷凝试验的试验本体结构，应用ANSYS软件对其结构进行了应力分析，并在现有结构的基础上对外部加强筋布置进行了一定的改进和优化，通过计算和比较可以看出，经过改进后的加强筋布置，不仅满足原有的试验要求，结构布置合理，更提高了试验本体的承压能力，使其能够满足更高试验压力的需要。
关键调：APWR;PCCS;压力容器；应力，加强筋
中图分类号：TL421;TL351
文章标志码：A
文章编号：0258-0918(2016)02-0159-06
Structureimprovementand analysisof pressureequipmentforUw
experimentaltestonsteamcondensationonthecoldsurface
YANG Lin
(State Nuclear Power Technology R&.D Centre, Beijing, 100190, China)
Abstract: The advanced pressurized water reactor (APWR) designed by Westinghouse uses a passive safety system which relies on heat removal by condensation to maintain the containment within the design limits of pressure and temperature. Steam condensation inside surface of the containment is one of the most important phenomena during heat removing process in the passive containment cooling system (PCCS). There was an experiment done by University of Wisconsin to study the mechanism of steam condensation on cold surface. In this paper, the structure of pressurized vessel of the test was introduced, and the pressure was calculated. Besides, the stiffener layout was improved. So the pressurized vessel can support higher pressure and also meet other thermal measurement requirements.
Keywords:APWR;PCCS;pressurizedvessel;stress;stiffener 修回日期：2015-10-19
基金名称：大型先进压水堆核电站重大专项，CAP1400非能动安全壳冷却系统性能研究及试验，2011ZX06002-005 作者简介：杨林（1981一)，男，宁夏银川，博士，工程师，核科学与技术专业，现主要从事反应堆安全壳研究
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