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首页> 外文期刊>Multidiscipline modeling in materials and structures >Conservative estimations of irradiation embrittlement of reactor pressure vessel steels for WWER-1000 lifetime prediction
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Conservative estimations of irradiation embrittlement of reactor pressure vessel steels for WWER-1000 lifetime prediction

机译：WWER-1000寿命预测的反应堆压力容器钢辐射脆化的保守估计

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摘要

Purpose - The purpose of this paper is to consider a procedure of water-water energetic reactor (WWER) reactor pressure vessel (RPV) lifetime prediction at the stages of design and lifetime extension using the standard irradiation embrittlement parameters as defined in regulatory documents. A comparison is made of the brittle fracture resistance (BFR) values evaluated using two criteria: shift in the critical brittleness temperature ΔT_c or shift in the brittle-to-ductile transition temperature ΔT_p and without shifts (T_c and T_p). Design/methodology/approach - The radiation resistance was determined using the following three approaches: calculation based on standard values ΔT_c and T_(c0) or ΔT_p and T_(p0) (a level of excessive conservatism); calculation based on standard value ΔT_c and actual value T_(c0) or actual values ΔT_p and T_(p0) (the level of realistic conservatism); or calculation based on actual values of T_c and T_(c0) or T_p and T_(p0) (the level of actual conservatism). The BFR was evaluated based on the results of testing the specimens subjected to irradiation in research reactors as well as surveillance specimens subjected to irradiation immediately under operating conditions. Findings - The excessive conservatism in determining the actual lifetime of nuclear reactor vessel materials can be eliminated by using the immediate values of critical brittleness temperature and ductile-to-brittle transition temperature. Originality/value - Obtained results can be applied to extend WWER vessel operating time at the stages of designing and operation due to substantiated decrease in conservatism. And it will allow carrying out a statistical substantiated assessment of the resistance to brittle fracture of the RPV steels.
机译：目的-本文的目的是考虑使用规范文件中定义的标准辐照脆化参数，在设计和寿命延长阶段对水-水能反应堆（WWER）反应堆压力容器（RPV）的寿命进行预测的程序。比较使用两个标准评估的脆性抗断裂性（BFR）值：临界脆性温度ΔT_c的变化或脆性至延性转变温度ΔT_p的变化且无变化（T_c和T_p）。设计/方法/方法-使用以下三种方法确定抗辐射性：基于标准值ΔT_c和T_（c0）或ΔT_p和T_（p0）（过度保守的程度）进行计算；根据标准值ΔT_c和实际值T_（c0）或实际值ΔT_p和T_（p0）（现实的保守程度）进行计算；或根据T_c和T_（c0）或T_p和T_（p0）的实际值（实际保守程度）进行计算。根据在研究反应堆中进行辐照的样品以及在运行条件下立即进行辐照的监测样品的测试结果，对BFR进行了评估。发现-通过使用临界脆性温度和延性至脆性转变温度的即时值，可以消除在确定核反应堆容器材料的实际寿命方面的过度保守性。原创性/价值-由于保守性的显着降低，获得的结果可用于在设计和操作阶段延长WWER容器的运行时间。它将允许对RPV钢的抗脆性断裂进行统计证实的评估。

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