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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脆化以支持寿命和寿命延长决策

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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:作为防止放射性出口的主要障碍,反应堆压力容器(RPV)是核电站(NPP)安全的关键组成部分。因此,减轻RPV在役脆性的所有可能措施必须满足当今对RPV可靠性提高的要求。已知退火处理是恢复由于中子辐照而劣化的RPV金属性能的有效措施。使用反应堆堆芯或一次回路泵可以在最高冷却液温度下进行低温“湿式”退火,尽管不能期望完全回收,但从实用的角度来看尤其如此,特别是在去除冷却液的情况下。内部是不可能的。通常,直到退火和辐照温度差达到70°C为止,都没有恢复作用。但是,众所周知,中子辐射与辐射脆化一起可以减轻金属中的辐射损伤。因此,我们试图检验利用核热利用手段将辐射诱导的渗碳作用在“湿式”退火技术中用作热和中子辐射源的可能性。为了支持上述构想,在15Cr3NiMoV型钢上进行了为期3年的反应堆实验,首先在运行中的加压水反应堆(PWR)中于270°C进行了初步辐照,随后在IR-8中进行了额外辐照(在330°C进行了87 h)测试反应堆已实现。实际上,脆化被部分抑制到高达中子注量密度降低1.5倍的值。在辐射增强退火的情况下,回收率等于27%,而炉退火在现有条件下的影响为零。提出了减轻辐射损伤的机理。希望“湿式”退火技术将有助于更好地管理RPV退化,这是影响核电厂寿命的一个因素,再加上相关的管理方法,将有助于促进压水堆的安全和经济的长期运行。

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