LOCA工况下锆合金包壳的行为概述

LOCA作为反应堆运行过程中比较严重的事故，是反应堆基准设计事故；而作为确保裂变产物不泄露的第一道屏障，锆合金优异的性能对于保障LOCA工况下的核安全具有重要意义。阐述了LOCA工况下锆合金的高温氧化行为、抗热冲击性能和力学性能及显微组织等方面的内容，为反应堆用锆合金的研发提供了技术支持。     

LOCA条件下包壳材料感应和电阻加热温升行为的对比研究

目的使有限元模拟技术成为一种切实有效的研究方法,进而为高性能反应堆包壳材料的设计以及可能发生的LOCA(Loss of Coolant Accident)事故下的应急措施等提供理论依据。方法基于COMSOL软件模拟分析典型锆合金核材料在LOCA条件下分别经感应加热和电阻加热后的温升行为。结果感应加热条件下,锆材的体积内最高温度、体积平均温度与表面中心点温度的差值随着温度上升逐渐增大,在1200℃瞬时温度下,温度差值最高,约为41℃。电阻加热条件下,锆材的体积内最高温度、体积内中心温度与表面中心点温度在加热的整个阶段近乎重合,最大差值约为3℃;锆材的体积平均温度、表面平均温度与表面中心点温度的差值出现负值,最大差值约为30℃。结论电阻加热和感应加热虽均适用于堆外研究反应堆失水事故下包壳材料所面临的超高温度及超快升温速率的工况模拟,但限于实际工况下电阻加热速率的滞后性,推荐使用感应加热进行后续的模拟研究工作。相关结果可为高性能反应堆包壳材料的设计提供必要的理论依据。     

锆合金涂层包壳专利申请态势全景分析

事故容错燃料(ATF)是日本福岛核事故之后提出的新一代核燃料概念,主要是为了提高反应堆在事故工况下的容错性能,从根本上提高核电厂对严重事故的抵御能力,从而有效地提高核电的安全性和经济性。针对传统核燃料使用的锆合金包壳,通过外表面涂层改性的方法提高其在事故工况下的抗高温氧化性能是事故容错燃料的主要研究方向。为了对锆合金涂层包壳的技术发展现状和趋势进行全面了解,对该技术相关的国际及国内专利申请态势进行了统计分析,研究了该领域的专利申请发展态势、专利来源地区及布局地区特点,重点分析了自20世纪60年代以来锆合金涂层包壳技术分支发展趋势;此外,结合我国的锆合金涂层包壳的研究现状,从研究进展及技术聚焦等方面提出锆合金涂层包壳研发建议。     

锆合金耐蚀性能影响因素概述

锆合金作为燃料包壳材料广泛应用于轻水反应堆。导致其在服役过程中腐蚀的因素复杂多样,随着合金成分、氧化物类型、第二相、晶粒形貌以及工作介质等的不同,其耐蚀性能会发生显著变化,分别介绍了这些因素对锆合金包壳腐蚀行为的影响。     

锆合金表面Cr基涂层的耐高温氧化性能

用磁控溅射法在锆合金基体表面制备Cr和CrAl层,并使其在1200℃/1 h水蒸汽中氧化,用扫描电子显微镜(SEM)、能谱仪(EDS)和X射线衍射仪(XRD)等手段表征氧化前后涂层和Zr合金基体的微观结构,研究了两种涂层在(反应堆失水(LOCA)事故情况下的)高温蒸汽环境中的抗氧化性能。结果表明：在1200℃/1 h水蒸汽中氧化后没有涂层的锆合金基体表面生成厚度约为100 μm的氧化膜;而在Cr涂层表面生成的致密Cr₂O₃层其厚度约为4 μm,表明氧化速率显著降低。CrAl涂层氧化后表面生成致密的Cr₂O₃和Al₂O₃混合氧化层,其厚度只有0.8 μm,表明氧化速率进一步降低。这些结果表明: 用磁控溅射法在锆合金表面制备的Cr和CrAl涂层,在1200℃水蒸气环境中均表现出良好的耐氧化性能。在Cr涂层表面生成的氧化膜厚度约为未涂层锆合金氧化层的1/25,CrAl涂层氧化膜厚度低于锆合金表面氧化层的1/100。     

典型事故容错轻水堆燃料包壳候选材料SiCf/SiC复合材料和Mo合金的研究进展

轻水堆是当前核电站应用最为广泛的堆型,其包壳材料均为锆合金。然而,福岛严重核事故的突发,使锆合金包壳的安全性受到质疑,事故容错燃料及其包壳候选材料被提上研究议程。本文综述了轻水堆用SiC_f/SiC复合材料和Mo合金包壳候选材料的研究进展,以及它们在轻水堆工况下的性能评估,指出实际工程应用所面临的挑战。最后展望了SiC_f/SiC复合材料和Mo合金在核燃料包壳中的应用前景。     

事故容错燃料包壳候选材料的研究现状及展望

2011年福岛核电站事故中,反应堆堆芯燃料中的锆合金包壳在事故工况下与高温水蒸汽发生剧烈氧化反应继而产生大量的氢气和热量,最终导致反应堆堆芯熔化和氢气爆炸,对社会和环境造成极大负面影响。自此之后,国内外纷纷展开对事故容错燃料的研究开发。相较于传统的UO2-Zr合金燃料体系,事故容错燃料能够在反应堆正常运行工况下维持或提高燃料性能,并在事故发生后相当长的一段时间内维持堆芯完整性,提供足够的时间裕量来采取事故应对措施。反应堆堆芯环境非常极端,包壳长期处于高温高压腐蚀介质中,同时还受到中子辐照的影响,因此新型包壳材料需要较好的耐腐蚀性和辐照稳定性。经不同研究者的研究评估,目前能够替代Zr合金的事故容错燃料包壳材料可分为陶瓷材料和金属材料两类:陶瓷材料主要以SiC/SiC复合材料为代表;金属材料主要有以FeCrAl为代表的Fe基合金和以Mo为代表的难熔金属及其合金。上述三种替代Zr包壳的材料各有其利弊,均未达到工程应用水平,并且都存在待解决的关键性问题。其中,FeCrAl合金的研发进展最快,目前在热学性能、力学性能、抗腐蚀性能、抗辐照性能等方面表现较好,但在工业加工和焊接等方面仍有待进一步改善。就SiC/SiC复合材料而言,由于SiC自身的高脆性而导致力学强度不足,不同的研究者提出了不同的结构设计思路试图降低包壳管失效概率,但包壳最终的结构设计仍未确定,而辐照引起的热导率急剧降低及连接密封和加工制造等方面还在不断研究中。Mo及Mo合金的力学性能和抗辐照性能较好,但自身抗腐蚀性较差,解决思路主要集中在提高钼纯度、调整合金的元素成分、进行表面涂层等方面。目前,对后两种材料包壳管的加工能力均未达到薄壁长管的工业制造水平。对于这几种候选包壳材料,需要建立属性数据库和一套完善的标准来衡量材料的质量。此外,还需开发相应的程序来评估包壳在堆内的行为。本文主要综述了SiC/SiC复合材料、FeCrAl合金、Mo及Mo合金三种候选包壳材料的研究进展,包括候选包壳材料的物理性质、耐腐蚀性能、力学性能、抗辐照性能、芯块-包壳力学与化学相互作用、在事故工况下的行为和工程应用等,综合分析了事故容错燃料包壳材料当前的研究现状,指出了各事故容错燃料包壳未来需集中解决的关键性问题。     

新一代锆锡合金

许多核反应堆制造商目前正考虑用锆和铌的二员与三元合金代替锆锡合金,用于轻水反应堆和重水反应堆。最近一些核反应堆卖主宣布已采用新一代合金-Zirlo合金,这是用铌改进的锆锡合金。这些材料可耐反应堆中的腐蚀和氧化,特别是在长时间的暴露情况,因而这些材料利于延长燃耗和使用长时间。     

锆-2合金的熔炼、加工和机械性能

从已发表的一些著作中可知,锆及钛合金的性能在很大程度上决定于很多因素,例如熔煉条件,成份,热处理等等。 本研究所持的目的系检验不同作者公布之锆-2合金来数据的再现性。 大家都知道,锆作为反应堆的结构材料是起着重要,作用的。例如用于非均匀反应堆中铀棒的把手,均匀反应堆中的铀的容器,均可以由锆制成。除对慢中子有良好的有效截面值之外(俘获截面为0.18±0.01巴恩/原子)     

锆材在核电站的应用及前景

锆合金因其优良的核性能和适宜的机械性能,在核电反应堆中作为包壳材料和结构材料得到了广泛应用.介绍了锆合金包壳管材在核电站中所起的重要作用,具体用途、用量,以及我国锆合金材料的研制和生产现状.指出随着我国核电事业的发展,对锆合金材料的需求仅更换的包壳管到2010年将达到120t,我国应加速核电用锆合金材料的国产化进程,这样才能自主地发展我国的核电事业.     

On the chemical compatibility between Zr-4 substrate and well-bonded Cr2AlC coating

Accident tolerant fuel(ATF) for the light water reactor has gained wide attentions after the Fukushima accident. To enhance the accident-tolerance of the nuclear system, one strategy is to modify the Zr-based alloy cladding surface with advanced ceramic coating. In this work, monolithic and dense Cr_2AlC coating has been synthesized by magnetron sputtering. The as-grown Cr_2AlC coating exhibits good chemical compatibility with Zr-based alloy substrate as well as mechanical integrity under both pull-off and scratch tests. The coating system also presents moderate thermochemical compatibility at 800℃ but degrades above 1000℃ under simulated loss-of-coolant accident(LOCA) conditions.     

容错事故燃料包壳用FeCrAl合金的研究进展EI北大核心CSCD

福岛事故后,人们迫切需要开发相应的燃料包壳材料以忍受严重事故发生时的极端工况,从而提高核电站的事故承受能力。尽管FeCrAl合金的宏观中子吸收截面要远远高于锆合金,但其在严重事故下良好的耐腐蚀性、优越的高温力学性能及抗辐照损伤能力,使其被列为事故容错燃料包壳的候选材料之一。然而,现有FeCrAl合金难以满足核电站用材料的要求,因此需对其进行优化,以获得更佳的性能。本文系统总结了近年来关于优化后FeCrAl合金的腐蚀行为、力学性能、辐照后的微观结构及力学性能变化、焊接性及加工性等方面的研究进展,分析了FeCrAl合金的高温腐蚀机理以及引起FeCrAl合金微观结构及力学性能变化的主要原因,提出了FeCrAl合金在高温腐蚀、焊接性以及加工性等过程中存在的主要问题以及未来的研究方向。     

A new approach to improving the macrosegregation in zirconium‐containing magnesium‐lithium alloy

In the preparation process of zirconium‐containing magnesium alloy, although zirconium is introduced into alloy by magnesium‐zirconium master alloy, the settling of zirconium particles has always been a key problem. In this study, the magnesium‐30wt.%zirconium master alloy was added into magnesium‐14wt.%lithium‐zinc alloy melt in the form of the block (about 20 mm) and the particles (about 20 μm), respectively, and then magnesium‐14wt.%lithium‐zinc with 0.5 wt.% zirconium alloy were prepared using stir‐casting process. Macrosegregation of major element (zirconium), microstructure and microhardness at different casting positions were examined to investigate the effect of zirconium addition methods on macrosegregation of magnesium‐14wt.%lithium‐zinc alloy. The results show that, for the block magnesium‐zirconium master alloy addition, there is obvious macrosegregation in alloy ingots, the zirconium contents at the top position of ingot are higher than that at the bottom by nearly 200 %. The method of particles master alloy addition can effectively improve macrosegregation, the difference in zirconium contents between the top and bottom is less than 16 %.     

Expert systems for emergency alarms analysis during accident situations in nuclear reactors

Expert system development is an important application of artificial intelligence. During the last few years, many successful expert systems have been developed in various fields like medical diagnosis, geological exploration, office management, etc. Expert systems is a computer software which solves reasonably complex problems where normally one needs an expert in that field to solve them.Recently efforts are being made in developing expert systems to support the operating staff in nuclear reactors. Nuclear power plant is one of the most complex engineering systems and safe and reliable operation is of primary importance. In spite of many automatic and redundant safety systems there are some occasions when the operating staff have to analyse the alarms and take further safety actions. Few of the severe accidents like Three Mile Island in USA and Chernobyl in USSR are attributed to the errors made by the operating staff and/or management. Nuclear engineers or systems analysts who may be expert in analysing an accident situation and advise corrective safety actions may not be readily available during the accident situations in the nuclear power plants. It is possible to model such expert knowledge in expert systems and this can be applied in diagnosing an accident situation like the loss of coolant accident (LOCA) and act as an additional confirmatory aid to the operating staff.Two small expert systems examples have been developed and are explained in this paper. One identifies a spurious LOCA alarm in a heavy water research reactor. The second example identifies the type of medium/small leakage (LOCA) in a coolant circuit of a PWR and suggests the subsequent safety actions. Both the examples have been developed using the expert systems shell VP-expert. They are off-line usable and user interactive. We do not propose expert system application for very fast response safety actions like reactor scram. These two small expert system examples are essentially to support the feasibility study in their applications during accident situations in nuclear power plants.     

Grain refining mechanism in Mg–9Gd–4Y alloys by zirconium

Abstract

The effect of different zirconium contents on the grain size of Mg–9Gd–4Y alloys and the grain refinement mechanism of zirconium have been studied. The results reveal that zirconium can refine the grains of the alloys to a large extent, and the grains become finer with an increase of zirconium content. Microstructural analysis shows that there is at least one zirconium rich core in almost each grain in alloys with a high zirconium content, whereas the characteristic zirconium rich cores are not found in the alloy with a low zirconium content. It is suggested that the grain refinement mechanism of zirconium in the low zirconium alloy is different from that in the high zirconium alloys: the zirconium works mainly by restricting grain growth in the low zirconium alloy, and by generating nucleants in the high zirconium alloys.     

Laser cladding of Zr-based coating on AZ91D magnesium alloy for improvement of wear and corrosion resistance

To improve the wear and corrosion resistance of AZ91D magnesium alloy, Zr-based coating made of Zr powder was fabricated on AZ91D magnesium alloy by laser cladding. The microstructure of the coating was characterized by XRD, SEM and TEM techniques. The wear resistance of the coating was evaluated under dry sliding wear test condition at room temperature. The corrosion resistance of the coating was tested in simulated body fluid. The results show that the coating mainly consists of Zr, zirconium oxides and Zr aluminides. The coating exhibits excellent wear resistance due to the high microhardness of the coating. The main wear mechanism of the coating and the AZ91D sample are different, the former is abrasive wear and the latter is adhesive wear. The coating compared to AZ91D magnesium alloy exhibits good corrosion resistance because of the good corrosion resistance of Zr, zirconium oxides and Zr aluminides in the coating.     

Ge、Si元素对ZrO2和Zr(Fe,Cr)2能量与电子结构的影响

采用基于密度泛函理论的第一性原理计算方法,研究Ge、Si元素对锆合金中与腐蚀相关的ZrO_2氧化膜相和Zr(Fe,Cr)_2第二相能量与电子结构的影响。合金形成热、结合能的计算结果表明:ZrO_2四方相结构不稳定,立方相易形成且结构稳定,氧化膜晶体结构从四方相向立方相发生转变影响锆合金的耐腐蚀性能;Ge、Si元素均降低ZrO_2立方相的结构稳定性和形成能力,与Ge相比,Si易取代Zr(Fe,Cr)_2第二相中的Cr,增加锆合金Fe/Cr原子比。电子态密度和Mulliken电子占据数的计算结果表明:ZrO_2中Zr与O存在杂化共振与较强的离子键作用,Ge、Si降低ZrO_2立方相结构稳定性的原因主要在于削弱了Zr-O之间的离子键作用;ZrO_2氧化膜相和Zr(Fe,Cr)_2第二相是影响锆合金耐腐蚀性能的两个重要因素,对Si而言,形成含Si的Zr(Fe,Cr)_2第二相对锆合金耐腐蚀性能产生不利影响,改善锆合金耐腐蚀性能需要ZrO_2晶体结构改变占主导地位;对Ge而言,含Ge的Zr(Fe,Cr)_2第二相难形成,第二相对锆合金耐腐蚀性能的影响相对Si较小,减缓ZrO_2由四方相向立方相的转变倾向,是Ge改善锆合金耐腐蚀性能的重要原因。