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

机译:退役锡安股份1核电站收获的RPV材料表征的现状

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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:伊利诺伊州锡安核电站1和2的退役,经过〜15〜15有效的全功率多年的服务,提供了独特的机会,以表征在职反应堆压力容器(RPV)材料的降解并评估目前可用的模型,用于预测RPV钢的辐射脆性[1-3]。此外,正在获得通壁厚度衰减和特性分布,并将结果与​​监控样本测试数据进行比较。预计这些努力将更好地了解与延长现有核电站(NPPS)的寿命超过60年的服务和随后许可更新的材料劣化。在支持美国核电反应堆舰队的扩展服务和当前运作,橡树岭国家实验室(ORNL),通过美国能源部,轻型水反应堆可持续发展(LWRS)计划,协调采购材料,组件等物品来自退役的锡安NPPS的兴趣。在本报告中,将讨论收获,切割样品块,加工试样,试验计划和从退役锡安NPP单元1中表征RPV的材料的现状。主要灶是圆周,焊粉80焊丝,线热72105(WF-70)带线焊缝和来自从峰值的区域收获的中间壳(0.7×10〜(19)n / cm〜2的中间壳体的A533B底座金属ZION单元1船舶内表面上的e> 1.0 mev)。在确定贯穿厚度的化学成分,夏比冲击,断裂韧性,拉伸和硬度测试,以表征基础金属和背带焊接材料的贯穿厚度力学性能。除了机械性能之外,使用各种微结构技术进行微观结构表征,包括原子探测层析造影,小角度散射和正电子湮灭光谱。

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