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Validation of Surveillance Concepts and Trend Curves by the Investigation of Decommissioned Reactor Pressure Vessels

机译:退役反应堆压力容器的调查验证监视概念和趋势曲线

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The investigation of reactor pressure vessel (RPV) material from the decommissioned Greifswald nuclear power plant representing the first generation of Russian-type WWER-440/V-230 reactors offers the opportunity to evaluate the real toughness response. This paper presents test results measured on trepans taken from the multilayer beltline welding seam SN0.1.4 and forged base metal ring 0.3.1 located in the reactor core region of the Unit 4 RPV. This unit was shut down after 11 years of operation. The characterization of the irradiation response was based on the measurement of the hardness, the tensile strength, the master curve reference temperature, T_0, and the Charpy-V transition temperature through the thickness of multilayer beltline welding seam SN0.1.4 and the forged base metal ring 0.3.1. For the beltline welding seam we observed a large variation in the through-thickness master curve T_0 values. The T_0 values measured with the T-S-oriented Charpy size SE(B) specimens strongly depended on the intrinsic welding bead microstructure along the crack tip. The progression of the T_0 values through the thickness-forged base metal ring 0.3.1 ranged from -121 to -130°C and indicated no irradiation-induced embrittlement within the fluence range of 5.38 to 1.20 x 10~(19) n/cm~2 (E > 0.5 MeV) through the thickness of the RPV wall. More than the allowed 2 % of the specimen size-adjusted fracture toughness values, K_(JC-1T), were below the fracture toughness curve for 2 % fracture probability. The reason for the occurrence of very low K_(JC-1T) values was seen in the intergranular planes detected on the fractured surfaces of the specimens. Modified master curve-based evaluation methods indicated the material to be nonhomogeneous. This investigation shows that the measured irradiation response of the investigated RPV materials does not correspond to the forecast according to the current Russian code.
机译:来自代表第一代俄罗斯型WWER-440 / V-230反应堆的退役Greifswald核电站反应堆压力容器(RPV)材料的研究提供了评估真正韧性反应的机会。本文介绍了从多层背带焊缝SN0.1.4和位于单元4RPV的反应器核心区域中的多层背带焊缝SN0.1.4和锻造基金金属环0.3.1上测量的测试结果。在11年运行后,本机已关闭。照射响应的表征基于硬度,拉伸强度,主曲线参考温度,T_0和夏比-V过渡温度的测量,通过多层背带线焊缝Sn0.1.4和锻造基础金属的厚度环0.3.1。对于带线焊缝,我们观察到贯穿厚度主曲线T_0值的大变化。用T-S取向夏比尺寸Se(B)测量的T_0值强烈地依赖于沿裂纹尖端的固有焊接珠粒微观结构。通过厚度锻造基础金属环的T_0值的进展0.3.1的范围为-121至-130℃,并在流量范围内没有显示在5.38至1.20×10〜(19)n / cm的情况下的辐照诱导脆化〜2(e> 0.5 meV)通过RPV壁的厚度。超过允许的2%的样品尺寸调节的断裂韧性值K_(JC-1T),低于断裂韧性曲线,裂缝率为2%裂缝概率。在样品的骨折表面上检测到的晶间平面中看到了非常低的K_(JC-1T)值的原因。修改的主曲线的评估方法表明了待非均匀的材料。本研究表明,根据当前俄罗斯代码,所研究的RPV材料的测量辐照反应与预测相对应。

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