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Etude des mécanismes de dégradation sous air à haute température des gaines de combustible nucléaire en alliage de zirconium

机译:Etudedesmécanismesdedégradationssousairàhautetempératuredesgaines decombustiblenucléaireenalliage de zirconium

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摘要

In nuclear plants, some accidental situations can result in air exposure of Pressurized Water Reactor (PWR) fuelassemblies: air ingress following a breach in the reactor vessel, deflooding during handling, spent fuel storagepool deflooding. Deprived of cooling source, the assemblies temperature raises and the fuel cladding, made out ofzirconium based alloys, oxidize. Compared to a steam oxidation, the degradation kinetic of the cladding is higher,on the one hand because of the high enthalpy of the zirconium-oxygen reaction (compared to zirconium-steamreaction), on the other hand because of the nitrogen contribution to the degradation. Temperature escalation andreaction runaway are expected and can rapidly lead to the loss of integrity of the cladding tubes.The objective of this PhD thesis was to affine the understanding of the high temperature air oxidation mechanismsof the two mostly used zirconium alloys in French PWR, Zircaloy-4 and M5®. Special attention has been paid toclarify the role of nitrogen.As-received Zircaloy-4 and M5® claddings segments have been oxidized in a thermobalance in air in isothermalconditions at temperatures between 800°C and 1000°C. Several characterization techniques (micro-Ramanspectroscopy, EPMA, XRD, optical and scanning electron microcopies...) have been used to analyze the oxidelayers. Identification and evolution of the different phases (monoclinic, tetragonal and cubic zirconia, zirconiumoxynitride and ZrN) has been evidenced and analyzed at several step of the oxidation process. Oxidationmechanisms have been proposed and the better oxidation resistance of the M5® alloy, compared to Zircaloy-4alloy, has been explained.The collected information will allow improvement of modeling aiming to predict the behavior of the claddings invarious accidental situations with air ingress (temperature transients, evolution of the gas phase composition…).
机译:在核电站中,某些意外情况可能导致压水堆(PWR)燃料组件暴露在空气中:反应堆容器破裂后空气进入,处理过程中发生驱油,乏燃料储存池驱油。缺乏冷却源,组件温度升高,由锆基合金制成的燃料包壳氧化。与蒸汽氧化相比,熔覆层的降解动力学更高,一方面是由于锆-氧反应的焓高(与锆-蒸汽反应相比),另一方面是由于氮对降解的贡献。预期温度升高和反应失控,它们会迅速导致覆层管的完整性丧失。本博士学位论文的目的是使人们对法国压水堆Zircaloy-2中最常用的两种锆合金的高温空气氧化机理有更深入的了解。 4和M5®。已特别注意阐明氮的作用。在空气中的热天平中,在等温条件下,温度介于800°C至1000°C之间,已氧化的Zircaloy-4和M5®覆层段已被氧化。几种表征技术(显微拉曼光谱,EPMA,XRD,光学和扫描电子显微复制……)已用于分析氧化层。在氧化过程的几个步骤中,已经证明并分析了不同相(单斜晶,四方晶和立方氧化锆,氮氧化锆和ZrN)的鉴定和演化。提出了氧化机理,并解释了与Zircaloy-4合金相比M5®合金具有更好的抗氧化性。所收集的信息将有助于改进建模,以预测随空气进入的各种偶然情况(温度瞬变)的包层行为。 ,气相组成的演变…)。

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    Idarraga Trujillo Isabel;

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  • 年度 2011
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  • 原文格式 PDF
  • 正文语种 fr
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