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首页> 外文期刊>Proceedings of the institution of mechanical engineers >Selection of candidate materials for reactor pressure vessels: Application of irradiation embrittlement prediction models and a stringency level methodology
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Selection of candidate materials for reactor pressure vessels: Application of irradiation embrittlement prediction models and a stringency level methodology

机译：反应堆压力容器的候选材料选择：照射脆化预测模型的应用和严格级方法

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摘要

The selection of materials for the reactor pressure vessel manufacturing is a complex process that involves great responsibility because small differences in chemical composition can adversely affect the manufacturing process and the in-service behavior of the material. Thus, it is recommendable to perform previous materials pre-selection stages based on the state-of-the-art knowledge, integrating research results with standardized requirements and using simplifier materials selection methodologies like the stringency level method. To address this issue, an evaluation of the influence of chemical composition on the shift of the ductile-to-brittle transition temperature has been performed using the most used and consolidated prediction models that are R.G. 1.99 Rev.2, NUREG/CR-6551, and ASTM E 900-02. A proposal of maximum limits for copper, nickel, and phosphorous to avoid irradiation embrittlement has been presented to carry out the process. The results have been analyzed by using the stringency level methodology to support the decision process. To this end, a materials data collection has been carried out to analyze the requirements described by 20 different specifications of materials from first to fourth generation of light water reactors, covering the main designs of pressurized reactors from Western Europe, North America, Japan, and Russia. It can be concluded that more recently developed materials exhibit more stringent requirements than earlier developed materials.
机译：用于反应器压力容器制造的材料是一种复杂的过程，涉及巨大的责任，因为化学组合物的小差异可能会对材料和材料的役性产生不利影响。因此，建议基于最先进的知识来执行以前的材料预选择阶段，将研究结果与标准化要求集成并使用简单的材料选择方法，如严格级别方法。为了解决这个问题，使用最常用的和综合预测模型进行了对延展性与脆性转变温度的影响的评估。 1.99 Rev.2，Nureg / CR-6551和ASTM E 900-02。已经提出了铜，镍和磷以避免辐射脆化的最大限制的提议以进行该过程。通过使用严格级别方法来分析结果来支持决策过程。为此，已经进行了一种材料数据收集，以分析来自第四代轻水反应堆的20种不同规格的材料所描述的要求，涵盖了来自西欧，北美，日本和的加压反应器的主要设计俄罗斯。可以得出结论，最近开发的材料比早期的开发材料更严格的要求表现出更严格的要求。

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