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CURRENT STATUS OF THE CHARACTERIZATION OF RPV MATERIALS HARVESTED FROM THE DECOMMISSIONED ZION UNIT 1 NUCLEAR POWER PLANT

机译:从退役的Zion装置1核电站收获的RPV材料的特性的最新状态

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The decommissioning of Units 1 and 2 of the Zion Nuclear Power Station in Zion, Illinois, after ~ 15 effective full-power years of service presents a unique opportunity to characterize the degradation of in-service reactor pressure vessel (RPV) materials and to assess currently available models for predicting radiation embrittlement of RPV steels [1-3]. Moreover, through-wall thickness attenuation and property distributions are being obtained and the results to be compared with surveillance specimen test data. It is anticipated that these efforts will provide a better understanding of materials degradation associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service and subsequent license renewal. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the U.S. Department of Energy, Light Water Reactor Sustainability (LWRS) Program, coordinated procurement of materials, components, and other items of interest from the decommissioned Zion NPPs. In this report, harvesting, cutting sample blocks, machining test specimens, test plans, and the current status of materials characterization of the RPV from the decommissioned Zion NPP Unit 1 will be discussed. The primary foci are the circumferential, Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B base metal from the intermediate shell harvested from a region of peak fluence (0.7 × 10~(19) n/cm~2, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following the determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing are being performed to characterize the through-thickness mechanical properties of base metal and beltline-weld materials. In addition to mechanical properties, microstructural characterizations are being performed using various microstructural techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy.
机译:伊利诺斯州锡安市锡安核电站1号和2号机组在约15年有效的全功率服务年后退役,这为表征在役反应堆压力容器(RPV)材料的退化和评估提供了独特的机会。目前可用于预测RPV钢的辐射脆化的模型[1-3]。而且,获得了穿墙厚度的衰减和特性分布,并将结果与​​监视样本测试数据进行比较。可以预期,这些努力将使人们更好地理解与将现有核电厂(NPP)的使用寿命延长至60年以上以及随后的许可证续期有关的材料退化。为了支持美国核反应堆舰队的扩展服务和当前运营,橡树岭国家实验室(ORNL)通过美国能源部的轻水反应堆可持续性(LWRS)计划,协调采购材料,组件和其他物品退役的锡安核电厂产生的利息。在本报告中,将讨论从退役的锡安核电厂1号机组中收获,切割样品块,加工测试样本,测试计划以及RPV材料表征的当前状态。主要焦点是周向,Linde 80焊剂,焊丝热72105(WF-70)束线焊缝和中间注量最大的区域(0.7×10〜(19)n / cm〜2)中收集的中间壳的A533B贱金属,E> 1.0 MeV)在Zion Unit 1容器的内表面。在确定贯穿厚度的化学成分之后,进行了夏比冲击,断裂韧性,拉伸和硬度测试,以表征贱金属和腰线焊接材料的贯穿厚度的机械性能。除机械性能外,还使用各种微结构技术进行微结构表征,包括原子探针断层扫描,小角中子散射和正电子hil没光谱。

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