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首页> 外文会议>ASME Pressure Vessels and Piping conference >APPLICATION OF SMALL PUNCH TESTING METHODS FOR THERMAL AGEING MONITORING AT PRIMARY CIRCUIT COMPONENTS IN NUCLEAR POWER PLANT
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APPLICATION OF SMALL PUNCH TESTING METHODS FOR THERMAL AGEING MONITORING AT PRIMARY CIRCUIT COMPONENTS IN NUCLEAR POWER PLANT

机译：小点测试方法在核电站主电路部件热老化监测中的应用

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This paper deals with the evaluation of material properties of safety-related components of the primary circuit of nuclear power plants (NPP) such as reactor pressure vessel (RPV), primary piping (PP) or steam generator (SG). The main degradation mechanism of NPP's components is radiation damage, but these processes are situated only in the core region of the reactor pressure vessel (RPV). The main mechanical loading of all individual parts of NPP's primary circuit are the influence of high pressure at elevated temperature till 300°C. These processes lead to the fatigue damage of structural materials, which is characterized as the thermal fatigue or thermal ageing. The reactors in Slovakia have been behind several decades in operation. Therefore, to ensure the safe and reliable operation of NPP, it is necessary to monitor and evaluate these changes. While the monitoring of the radiation degradation is standardly performed using surveillance programs during all plant life operation, the monitoring of thermal ageing at primary circuit components was realized only after few years of all NPP units operation. The concept of the thermal ageing assessment was divided into two possible approaches to evaluation. The first is long-term exposition of surveillance specimens. And the second approach is a direct surface sampling of heavy components after several years of operation. The Small Punch Test (SPT) methods are mainly used for evaluation of materials actual state. By the SPT technique it is possible to evaluate the basic tensile properties as the ultimate tensile strength and the yield stress of the tested materials from a very small amount of obtained material. The details of the original VUJE design program of thermal ageing monitoring of NPP primary circuit materials in Slovakia is described in this paper.
机译：本文涉及对核电厂一次回路安全相关组件（例如反应堆压力容器（RPV），一次管路（PP）或蒸汽发生器（SG））的材料性能的评估。 NPP组件的主要降解机理是辐射损伤，但这些过程仅位于反应堆压力容器（RPV）的核心区域。 NPP一次回路的所有各个部分的主要机械负载是高温直至300°C的高压影响。这些过程导致结构材料的疲劳损伤，其特征是热疲劳或热老化。斯洛伐克的反应堆已经落后了几十年。因此，为确保核电厂安全可靠运行，有必要对这些变化进行监测和评估。虽然通常在所有工厂生命周期运行期间都使用监视程序对辐射退化进行监视，但仅在所有NPP单元运行几年后，才实现对一次回路组件热老化的监视。热老化评估的概念分为两​​种可能的评估方法。首先是监视标本的长期暴露。第二种方法是经过几年的运行直接对重组分进行表面采样。小打孔测试（SPT）方法主要用于评估材料的实际状态。通过SPT技术，可以从非常少量的获得的材料中评估基本拉伸性能，即极限拉伸强度和被测材料的屈服应力。本文介绍了原始的VUJE设计程序的详细信息，该程序在斯洛伐克对NPP一次回路材料进行热老化监测。

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