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Controlling RPV Embrittlement Through Wet Annealing In Support of Life Attainment and Life Extension Decisions

机译:通过湿退火控制RPV脆化,以支持生命达到和生命延伸决策

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As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature ?wet? annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70°C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in ?wet? annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270°C and following extra irradiation (87 h at 330°C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that ?wet? annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated management methods, will help facilitate safe and economic long-term operation of PWRs.
机译:作为放射性出口反应器压力容器(RPV)的主要屏障是核电厂(NPP)安全性的关键组成部分。因此,RPV可靠性增强的当今需求必须通过用于RPV在职脆化缓解的所有可能的动作。已知退火处理是恢复通过中子辐射劣化的RPV金属性能的有效措施。低温?湿?在最大冷却剂温度下,可以使用反应器芯或初级回路泵获得的最大冷却剂温度,尽管不能预期产生完全恢复,从实际的观点中更具吸引力,特别是在案件中是不可能的。通常,没有恢复效应,退火和70°C的辐射温度差。然而,已知的是,随着辐射脆化中子辐射以及可以减轻金属的辐射损伤。因此,我们试图测试使用辐射诱导的液位效果的可能性?潮湿?通过核热利用作为热量和中子照射来源的退火技术。为了支持上述概念,在15Cr3Nimov型钢上具有3年的持续时间反应器实验,在270℃下使用加压水反应器(PWR)和IR-8在330℃下的额外照射(87小时)下进行初步辐射满足测试反应器。实际上,脆化部分抑制了相当于1,5倍中子流量的值。辐射增强退火的情况下的恢复程度等于27%而炉退火导致在现有条件下零效应。提出了辐射诱导损伤的机制。希望这是湿吗?退火技术将有助于提供RPV降解的更好管理作为影响核电站寿命的因素,其中与相关的管理方法一起有助于促进PWR的安全和经济的长期运行。

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