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基于卡尔曼滤波的热老化性能可靠性预测

Performance Reliability Prediction for Thermal Aging Based on Kalman Filtering
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摘要 以核电站主管道为研究对象,运用性能退化可靠性理论,对主管道的热老化性能可靠性进行了研究。首先通过加速热老化实验获得的数据,分析主管道奥氏体不锈钢材料冲击性能及断裂韧性的退化过程,利用状态空间方法建立了时变性能退化量模型,并通过卡尔曼滤波对性能趋势进行预测;然后考虑冲击性能与断裂韧性之间的相关性,运用随机过程理论建立了基于多性能参数的主管道热老化实时性能可靠性预测模型,从而得到多参数下的主管道热老化性能可靠度及可靠性寿命,为核电站进行主管道老化维修决策优化管理提供了科学依据。 The performance reliability of the nuclear power plant main pipeline that failed due to thermal aging was studied by the performance degradation theory .Firstly , through the data obtained from the accelerated thermal aging experiments ,the degrada‐tion process of the impact strength and fracture toughness of austenitic stainless steel material of the main pipeline was analyzed .The time‐varying performance degradation model based on the state space method was built ,and the performance trends were pre‐dicted by using Kalman filtering .T hen ,the multi‐parameter and real‐time performance reliability prediction model for the main pipeline thermal aging was developed by consid‐ering the correlation between the impact properties and fracture toughness ,and by using the stochastic process theory .T hus ,the thermal aging performance reliability and relia‐bility life of the main pipeline with multi‐parameter were obtained ,which provides the scientific basis for the optimization management of the aging maintenance decision mak‐ing for nuclear power plant main pipelines .
出处 《原子能科学技术》 EI CAS CSCD 北大核心 2015年第5期909-914,共6页 Atomic Energy Science and Technology
基金 国家科技重大专项资助项目(2011ZX06004-002) 国家自然科学基金资助项目(51105344) 航空科学基金资助项目(2102ZB55003) 河南省基础与前沿技术研究基金资助项目(132300410269) 河南省教育厅科学技术重点研究项目资助(14A590001)
关键词 主管道 核电站 热老化 性能衰退 卡尔曼滤波 状态空间方法 main pipeline nuclear power plant thermal aging performance degrada-tion Kalman filtering state space method
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  • 1刘鹏,薛飞,戴忠华,陈世均,朱文彬,汪小龙,遆文新.轻水堆核电站奥氏体不锈钢铸件的热老化及其老化管理[J].核动力工程,2005,26(S1):93-96. 被引量:24
  • 2束国刚,陆念文.压水堆核电厂关键金属部件的老化和寿命评估[J].中国电力,2006,39(5):53-58. 被引量:22
  • 3李颖,刘涛,栾培锋,史巨元.核电厂压水堆主管道材料性能的研究[J].物理测试,2006,24(5):12-13. 被引量:5
  • 4赵彦芬,遆文新,汪小龙,薛飞.核电站用钢管材料及其国产化[J].钢管,2007,36(2):11-14. 被引量:27
  • 5SAMUEL K G. Evaluation of Ageing-induced Embrittlement in an Austenitic Stainless Steel by Instrumented Impact Testing[J]. Journal of Nuclear Materials, 1987,150 : 78.
  • 6Chung H M. Aging and Life Prediction of Cast Duplex Stainless Steel Components[J]. Int J Press Vessels Piping. 1992,50 : 179.
  • 7Iacoviello F, Casari F, Gialanella S. Effect of "475 ℃ Embrittlement" on Duplex Stainless Steels Localized Corrosion Resistance[J]. Corrosion Science, 2005,47 : 909.
  • 8French Association for Design, Construction and Inservice Inspection Rules for Nuclear Island Components. RCC-M-2000, Design and Construction Rules for Mechanical Components of PWR Nuclear Islands [S]. Paris : AFNOR, 2002.
  • 9Chopra O K, Sather A. Initial Assessment of the Mechanisms and Significance of Low Temperature Embrittlement of Cast Stainless Steels in LWR Systems. US Nuclear Regulatory Commission Report: NUREG/CR-5385, Argonne National Laboratory, 1990.
  • 10Chopra O K, Clung H M. Aging of Cast Duplex Stainless Steels in LWR Systems[J]. Nucl Eng Des, 1985,89 305.

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