共查询到18条相似文献,搜索用时 500 毫秒
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反应堆压力容器是核电厂最重要设备之一,其辐照脆化状态决定了核电厂的实际运行寿命。通过借鉴国外反应堆压力容器安全评估方法,开发出一套反应堆压力容器辐照脆化时限老化分析(TLAA)的方法。该方法从上平台能量、反应堆运行压力-温度曲线及承压热冲击3个方面评价压力容器材料在正常工况和事故工况下的安全裕度。采用该方法在秦山核电厂运行许可证延续(OLE)项目中对反应堆压力容器进行了辐照脆化TLAA安全评估,其评估方法和评估结论到得国家核安全监管局的认可,为秦山核电厂延寿20 a奠定了基础。 相似文献
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核电厂运行许可证延续必须考虑其延寿期内的核安全问题,确保核电机组在延期运行期间的核安全水平不低于原设计寿期内的核安全水平。可应用PSA技术对许可证延续期间的核电厂建立老化PSA模型,从而评估SSC老化对核电厂整体安全的影响,验证其仍可满足原设计标准。基于此提出了应用于核电厂老化PSA的SSC筛选分析方法,通过考虑趋势分析,老化失效模式与影响分析,风险重要度分析,在三种分析方法基础上建立核电厂SSC筛选的决策矩阵,为选择易老化且安全重要的部件建立了可行的方法。该项工作也为核电厂在许可证延续阶段的风险指引型管理奠定技术基础。 相似文献
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焦丽玲 《核标准计量与质量》2012,(4):26-31
通过对国内外核电厂老化管理经验的调研,以核电厂机械设备老化管理为主要关注点,结合国内老化管理经验和IAEA的老化管理指导原则,从老化管理的方法和步骤、老化管理设备筛选原则及分级、老化机理研究、系统化老化管理的模式和方法四个方面切入分析,提出了核电厂老化管理大纲内容的编写思路和建议。 相似文献
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In less than 10 years, the first commercial pressurized water reactor (PWR) plant in Korea will reach its official design life. As part of safety activities, developed countries have already implemented periodic safety review (PSR) or equivalent programs to check and improve the safety of operating nuclear power plants (NPP) during their plant life. At the end of 1999, it was decided by the Korean Atomic Energy Safety Committee to adopt the PSR program and to apply it to Korean operating NPP. Since Kori Unit 1 started the review for the first tentative application of PSR as a model case in May 2000, it is now progressing well. Management of aging is one of the major factors to be considered in PSR and life extension of a nuclear plant. This paper is intended to introduce the regulatory aspect and strategy of Korean PSR. The background and scope of basic PSR guidelines are described, and a summary of technical criteria for aging management, which shows a regulatory direction for PSR, is also presented. 相似文献
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恰希玛核电厂压力容器各关键部位在PTS瞬态下的温度场研究 总被引:1,自引:1,他引:0
承压热冲击现象在核电厂延寿评估中应被重点关注。本文针对恰希玛核电厂1号机组的压力容器及堆内构件建立了完整的CFD模型,计算了正常工况下压力容器内冷却剂的速度场和温度场分布,计算结果与试验结果符合良好。本文详细研究了蒸汽发生器传热管破裂事故工况下压力容器接管及下降段中冷却剂的热工水力特性,并将计算结果与RELAP5计算结果进行对比,结果表明二者符合良好。本文研究可为反应堆压力容器老化管理评估的计算分析工作提供重要参考。 相似文献
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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。 相似文献
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The adequate margin of safety reached in the as-built condition for a nuclear power plant shall be maintained throughout the whole life of the plant. To attain this, a systematic lifetime evaluation of safety related items should be performed in due time in the light of new developments. The results can be used for the purpose of life extension and license renewal, too. Siemens has gained an integral analysis concept by practical experience. The analysis can result in corrective actions with respect to life extension of plant components and systems or can lead to measures for plant improvement. Examples of performed activities are given within this article. 相似文献
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生态环境部第8号令《核动力厂、研究堆和核燃料循环设施安全许可程序规定》对核动力厂、研究堆和核燃料循环设施运行许可证件延续事项作出了新的规定。为推动我国研究堆老化管理标准体系建立,分析了我国研究堆延寿审查策略发展历程,结合高通量工程试验堆等研究堆运行许可证有效期延续申请审查工作中的几个关键问题,提出了以定期安全审查为主、重点依据老化管理并兼顾技术规格书审查及差异性审查的审查策略,研究成果为我国研究堆老化管理法规标准的建立提供了实践经验及理论指导依据。 相似文献
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D.J. Naus C.B. Oland B.R. Ellingwood H.L. Graves III W.E. Norris 《Nuclear Engineering and Design》1996,166(3):453
Research is being conducted by Oak Ridge National Laboratory under US Nuclear Regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants. 相似文献