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Long Term Irradiation Effects on the Mechanical Properties of Reactor Pressure Vessel Steels from Two Commercial PWR Plants

机译:两种商用PWR植物反应堆压力容器钢的力学性能的长期辐照效应

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The Swedish nuclear power plants all have plant specific surveillance programs which includes samples from all relevant materials that are subjected to a fluence-level that exceeds 1~*10~(17) n/cm~2 over the estimated period of operation for the specific power plants. The Swedish pressurized water reactor (PWR)-plants are currently planning for a service period beyond 50 years of operation. As a portion of that, two of the three PWR units at the Ringhals site are conducting a major effort to verify the fitness to service of the reactor pressure vessel (RPV). In this case it is the weld in the belt-line region of the RPV, which is the apparent limiting factor. The weld metal contains high Nickel and high Manganese levels, not commonly used in other PWR-reactors. The effort includes a densified testing of the available surveillance capsule material in order to better understand the degradation phenomena and also an extended testing scope. A spin off effect of this program is that high fluence data for the base material also is made available from the testing. The chemical composition of the base metal is valid for many of the currently operating PWR-vessels. This study is an analysis of both the weld and the base material data extracted from the surveillance program. The results are evaluated against currently available data and correlation curves. In general, the results point out that the current Regulatory Guide 1.99 revision 2-correlation regarding the prediction of as-irradiated transition temperature is under-conservative for the tested material. The transition temperature shift, here evaluated as the temperature shift at 41J, is under-predicted by the correlation by as much as 70°C in some cases and increases with increasing fluences. However, prediction made by the French average irradiation embrittlement prediction formula, FIM-formula, is consistently better but still slightly under conservative.
机译:瑞典核电厂都具有植物特异性监测程序,包括来自所有相关材料的样品,该样品受到超过1〜* 10〜(17)n / cm〜2的流量水平,在估计的特定运营期间发电厂。瑞典加压水反应器(PWR)植物目前正在计划运行50年以上的服务期。作为其中的一部分,Ringhals现场的三个PWR单元中的两个是进行重大努力,以验证反应器压力容器(RPV)的适应性。在这种情况下,它是RPV的带线区域中的焊缝,这是表观限制因子。焊接金属含有高镍和高锰水平,不常用于其他PWR-反应器。努力包括对可用的监视胶囊材料的致密测试,以便更好地理解降级现象以及扩展测试范围。该程序的旋转效果是基础材料的高通量数据也可从测试中获得。基础金属的化学成分对于许多目前操作的PWR血管有效。本研究是对从监视程序提取的焊缝和基础材料数据的分析。结果针对目前可用的数据和相关曲线进行评估。通常,结果指出,当前调节指南1.99修复2-相关性对测试材料的预测是辐照温度的预测。这里的转变温度转变为41J的温度偏移,在某些情况下通过多达70℃的相关性被相关,并且随着流量的增加而增加。然而,由法国平均照射脆化预测公式,FIM配方制成的预测始终如一,但仍然略微在保守范围内。

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