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热蠕变对UMo/Zr单片式燃料板起泡行为的影响 被引量:1

Effects of Thermal Creep on Blister Behavior in UMo/Zr Monolithic Fuel Plates
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摘要 针对含有气腔的UMo/Zr单片式燃料板,考虑包壳材料的热蠕变效应,将包壳的变形与气腔压力相耦合,发展了一种对燃料板宏观起泡行为进行数值模拟的方法。基于所建立的模拟方法,计算分析了包壳热蠕变和气腔内裂变气体原子数对起泡行为的影响。研究发现,在考虑包壳热蠕变时,若局部开裂区域内的裂变气体原子数为4.0×1017,以鼓泡高度0.1 mm作为起泡阈值的判断标准,所预测出的阈值温度比不考虑热蠕变时低100℃;若局部开裂区内的裂变气体原子数由2.5×1017增加至4.0×1017,则燃料板的起泡阈值温度将可能降低40℃,通过降低包壳材料的热蠕变率可以有效提高燃料板的抗鼓泡能力。 For a UMo/Zr monolithic fuel plate with a gas space,a method is developed to simulate the macroscale blister behavior considering the thermal creep effects of the cladding,in which the calculation of cladding deformation is coupled with the gas space pressure.Based on the developed simulation method,the effects of thermal creep strain of cladding and the internal fission gas atom number on the blister behavior are analyzed.The research results indicate that with the thermal creep of cladding considered,if the fission gas atom number is 4.0×1017,the predicted blister threshold temperature will be 100℃lower than the case without considering the thermal creep of cladding,with the blister threshold temperature set as the temperature at which the blister height reaches 0.1 mm,with the fission gas atom number increasing from 2.5×1017 to 4.0×1017,the blister threshold temperature might decrease by 40℃.The blister threshold temperature of the fuel plates could be improved by using a cladding material with low thermal creep rate.
作者 严峰 简晓彬 丁淑蓉 辛勇 唐昌兵 李垣明 Yan Feng;Jian Xiaobin;Ding Shurong;Xin Yong;Tang Changbing;Li Yuanming(Department of Aeronautics and Astronautics,Fudan University,Shanghai,200433,China;Science and Technology on Reactor System Design Technology Laboratory,Nuclear Power Institute of China,Chengdu,610213,China)
出处 《核动力工程》 EI CAS CSCD 北大核心 2020年第6期85-91,共7页 Nuclear Power Engineering
基金 国家自然科学基金资助项目(11572091,11772095) 国家重点研发计划资助项目(2016YFB0700103) 中国核动力研究设计院核反应堆系统设计技术重点实验室资助项目(HT-LW1Y-09-2017010)。
关键词 UMo/Zr单片式燃料板 起泡阈值温度 退火试验 热蠕变 数值模拟 UMo/Zr monolithic fuel plates Blister threshold temperature Blister anneal test Thermal creep Numerical simulation
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  • 1丁淑蓉,霍永忠.弥散型燃料板辐照肿胀行为的有限元分析[J].核动力工程,2007,28(4):58-62. 被引量:3
  • 2HOLDEN A N. Dispersion fuel elements[M]. New York: Gordon and Breach, Science Publish ers, Inc. , 1967: 3-13.
  • 3ADEI.FANG P, RITCHIE I O. Overview of the status of research reactors worldwide[C] // 2003 International RERTR Meeting. Chicago, Illi nois: Argonne National Laboratory, 2003.
  • 4DIENST W, NAZARE S, THUMMLER F. Irradiation behavior of UAI,-A1 dispersion fuels for thermal high flux reactors[J]. Journal of Nuclear Materials, 1977, 64(1): 1-13.
  • 5IAEA. Good practices for qualification of high density low enriched uranium research reactor fuels[R]. Vienna: IAEA, 2009.
  • 6SNELGROVE J L, HOFMAN G L. Evaluation of existing technology base for candidate fuelsg for the HWR NPR [R]. USA: Argonne National Laboratory, 1993.
  • 7WEIR J R. A failure analysis for the low-temper- ature performance of dispersion fuel elements[R]. USA: Oak Ridge National Laboratory, 1960. BECK S D. Failure analysis of dispersion fuel elements based on matrix cracking ER2. USA:ALCO Products, Inc. , 1961.
  • 8BECK S D. Failure analysis of dispersion fuel elements based on matrix cracking [R]. USA: ALCO Products, Inc. , 1961.
  • 9WILDER A S, KELLEMAN R W. Determina- tion of matrix permissible irradiation on type 3 (SM-2) fuel plates[R]. USA: ALCO Products, Inc. , 1962.
  • 10KELLER D L. Predicting burnup of stainless UOe cermet fuels[J]. Nucleonics, 1961, 19(6): 45-48.

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