High-Temperature Strength of KTTs-110 Zirconium Alloy

We describe the methods and results of investigation of the characteristics of strength and plasticity of a pilot batch of tubes made of KTTs-110 zirconium alloy under the conditions of short- and long-term static loading. The obtained results are compared with the same characteristics of É-110 alloy extensively used in the active zones of WWÉR-1000-type nuclear reactors.     

锆合金的低周疲劳特性

锆合金被普遍用做核反应堆中的燃料包壳和结构材料。在反应堆运行时，堆功率的波动和水冷却介质的流动．使燃料组件及其它构件发生循环变形，在极端情况下出现破损。本文概述了堆内包壳循环变形的特点，并分析了锆合金的循环变形行为，疲劳裂纹的形核与扩展，疲劳寿命及影响疲劳寿命的因素。     

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...     

Zirconium alloys in nuclear technology

This paper describes the historical development of zirconium and its alloys as structural materials for nuclear reactors. The various problems encountered in the early stages of the development of zircaloys and their performance in reactors operating presently are described in detail. The development of Zr-2.5 % Nb alloys for pressure tube applications is discussed. The paper concludes with a detailed discussion on the development potential of zirconium alloys for high temperature applications and a brief account of the work carried out at Trombay in this field.     

不同辐照粒子下钨及钨合金辐照损伤行为的研究进展

研究核聚变、准稳态等离子体下面向等离子体材料的辐照行为,发展适合于先进实验超导托卡马克(EAST)、国际热核聚变实验堆(ITER)和中国聚变工程实验堆(CFETR)长脉冲高参数运行乃至未来聚变反应堆稳态运行的高性能面向等离子体材料是当前核聚变研究一项艰巨而又紧迫的任务。钨因具有高熔点、高导热率、低溅射腐蚀速率、高自溅射阀值以及低蒸气压和低氚滞留等优异性能,被认为是聚变装置最具有前景的面向等离子体材料。综合评述了钨及钨合金在不同辐照粒子下损伤行为的最新研究进展。粒子辐照造成的微观缺陷在钨及钨合金内部累积,辐照造成缺陷的形成和数量与钨基材料颗粒微观结构、第二相成分等密切相关,辐照缺陷情况各异。同时,辐照粒子种类、能量、剂量和温度等辐照条件都会对钨材料辐照后的形貌特征和缺陷产生重要影响。     

轻水反应堆（LWR）用包壳材料研究进展

轻水反应堆（LWR）是国际上多数核电站采用的堆型。锆具有良好的加工性能,优良的机械性能,较高的熔点、优异的耐蚀性能及核性能,被用作燃料包壳和堆芯结构材料,是发展核电及核动力舰船不可替代的关键结构材料和功能材料。随着核电技术的发展,对堆芯包壳材料性能提出了更高的要求,综述了核用锆合金包壳材料的国内外研究和使用现状以及新型SiC包壳材料的研发现状。总体来说,锆合金在未来几十年内仍是核反应堆包壳材料的主要用材,开展新合金的研发,不断提升锆合金的性能是世界各国研究者共同的目标;适时加大投入力度,强化条件建设,就能加快具有国内自主知识产权锆合金的产业化步伐,可最终实现核电及核动力用锆合金材料的自主化;SiC材料具有更高的熔点、更好的耐腐蚀性能,是一种极具应用潜力的材料,有可能成为第4代核反应堆的包壳材料,但还需投入大量研究。     

Crystal Structure of the ZrO Phase at Zirconium/Zirconium Oxide Interfaces

Zirconium‐based alloys are used in water‐cooled nuclear reactors for both nuclear fuel cladding and structural components. Under this harsh environment, the main factor limiting the service life of zirconium cladding, and hence fuel burn‐up efficiency, is water corrosion. This oxidation process has recently been linked to the presence of a sub‐oxide phase with well‐defined composition but unknown structure at the metal–oxide interface. In this paper, the combination of first‐principles materials modeling and high‐resolution electron microscopy is used to identify the structure of this sub‐oxide phase, bringing us a step closer to developing strategies to mitigate aqueous oxidation in Zr alloys and prolong the operational lifetime of commercial fuel cladding alloys.     

Remote Field Eddy Current Control Using Rotating Magnetic Field Transducer: Application to Pressure Tubes Examination

The remote field eddy current (RFEC) technique is used to investigate the possibility of detecting the discontinuities practiced on pressure tubes samples from nuclear reactors, pressurized heavy water reactors (PHWR) type. In this article, we propose to develop the RFEC using the technique of rotating magnetic field (RMF). A method for calculating the field generated by the eddy current transducer with RMF using propagator matrix was developed. The experimental measurements are realized for artificial discontinuities practiced in pressure tubes samples.     

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。     

Fatigue testing of materials under extremal conditions by acoustic method

Increasing fuel cycle time requires fatigue testing of the fuel clad materials for nuclear reactors. The standard high-temperature fatigue tests are complicated and tedious. Solving this task is facilitated by the proposed acoustic method, which ensures observation of the material damage dynamics, monitoring of the experimental parameters, and determination of the dynamic yield stress. Ring samples cut from zirconium cladding tubes were irradiated with fast neutrons to a total fluence of 2.2×10²⁶ m^?2 and fatigue tested at temperatures up to 360–400°C, including the tests in an iodine vapor—a uranium fission product capable of inducing corrosion cracking of the fuel elements.     

Limit load analyses of helical coiled steam generator tubes with a volumetric flaw

Maintaining structural integrity of piping has always been an important effort in the nuclear power industry. To resolve piping issues such as unanticipated failures caused by diverse planar and volumetric flaws, several guidelines were developed and used especially for assessment of steam generator tubes. However, because the major components of new reactors have dissimilar geometric features and loading conditions compared with those of conventional operating reactors, most of existing assessment methods are not expected to be applicable; therefore, new alternative assessment guidelines are required. In this paper, a systematic structural integrity assessment of helical coiled steam generator tubes for a small and medium modular reactor, which is currently being designed, is introduced. Three‐dimensional detailed finite element (FE) limit analyses have been carried out to simulate the behaviours of the tube containing a volumetric flaw such as elliptical wear‐type, rectangular wear‐type and tapered wear‐type defects subjected to external pressure. Failure pressures were calculated from the FE analyses by changing defect depth, defect length and defect angle affecting the load‐carrying capacity of the tube. Thereby, engineering equations were developed as a function of these key parameters to predict structural failures and system reliability to enable more reasonable design and manufacturing decisions.     

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。     

B30铜镍合金换热管腐蚀与防护研究进展

B30铜镍合金具有优良的耐海水腐蚀性能,一般用作船舶海水冷凝器系统的换热管路,但在实际工况中仍因各种原因存在局部腐蚀失效问题.论述了B30铜镍合金海水换热管路的腐蚀问题、腐蚀机理、腐蚀影响因素及其防腐措施,对合理设计船舶冷凝器换热管路防腐蚀设计具有一定的参考意义.     

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。     

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。     

Chlorination process applied to zirconium recovery from Zircaloy shavings

Abstract

The chlorination process is an attractive method for recovering metals from complex scrap. The process, which is performed at high temperatures, is able to convert all metals contained in the scrap into their respective chlorides. Then, they can be separated on the basis of their different physical properties. Finally, they can be reduced to metals by the Kroll process. This paper presents the application of the chlorination process to recover zirconium from scrap Zircaloy, which is a very common zirconium based alloy used in the nuclear industry. The study was carried out using thermogravimetric analysis, scanning electron microscopy, energy dispersive X-ray spectroscopy, and X-ray diffraction analysis.     

A review on neutron-irradiation-induced hardening of metallic components

Neutron irradiation alters the mechanical properties of metallic parts, which are exposed to service temperatures below 40% of their homologous temperature. These working conditions affect most of the components of fission nuclear reactors, making these parts susceptible during service to hardening, loss of ductility, localised plastic deformation and plastic instability. Additionally, there has been a continuous historical increase in the efficiency and service life of nuclear reactors, leading to more severe irradiation exposure during service. In this sense, understanding the mechanisms for the formation and evolution of irradiation-induced defects and their interaction with gliding dislocations is vital for the estimation of the service life of these components and the development of new radiation-resistant materials via alloy and microstructural design. The present paper reviews the use of atomic-scale modelling to simulate the generation and evolution of irradiation-induced defects. Additionally, the interaction between these defects and the gliding dislocations is revised in accordance with the continuum theory and atomic-scale modelling. Finally, the limitations and challenges facing the atomic-scale modelling of radiation damage and defect/dislocation interaction are briefly discussed.     

Numerical simulation of hydrogen evolution upon interaction between zirconium melt and water

The process of hydrogen generation upon interaction between a zirconium melt and water is studied. An oxidation model for zirconium liquid droplets in a water–steam medium was developed. The model was incorporated into the VAPEX computational code developed for calculation of interactions between core material melts and water during a severe accident at nuclear power plants equipped with pressurized water reactors. The modified VAPEX code was used to analyze the ZREX experiments on molten zirconium–water interactions. It was shown that the VAPEX code satisfactorily predicts the amount of the hydrogen generated during such an interaction.     

Optimization of N18 Zirconium Alloy for Fuel Cladding of Water Reactors

In order to optimize the microstructure and composition of N18 zirconium alloy(Zr-1Sn-0.35Nb-0.35Fe-0.1Cr,in mass fraction,%),which was developed in China in 1990s,the effect of microstructure and composition variation on the corrosion resistance of the N18 alloy has been investigated.The autoclave corrosion tests were carried out in super heated steam at 400 ℃/10.3 MPa,in deionized water or lithiated water with 0.01 mol/L LiOH at 360 ℃/18.6 MPa.The exposure time lasted for 300-550 days according to the test temperature.The results show that the microstructure with a fine and uniform distribution of second phase particles(SPPs),and the decrease of Sn content from 1%(in mass fraction,the same as follows) to 0.8% are of benefit to improving the corrosion resistance;It is detrimental to the corrosion resistance if no Cr addition.The addition of Nb content with upper limit(0.35%) is beneficial to improving the corrosion resistance.The addition of Cu less than 0.1% shows no remarkable influence upon the corrosion resistance for N18 alloy.Comparing the corrosion resistance of the optimized N18 with other commercial zirconium alloys,such as Zircaloy-4,ZIRLO,E635 and E110,the former shows superior corrosion resistance in all autoclave testing conditions mentioned above.Although the data of the corrosion resistance as fuel cladding for high burn-up has not been obtained yet,it is believed that the optimized N18 alloy is promising for the candidate of fuel cladding materials as high burn-up fuel assemblies.Based on the theory that the microstructural evolution of oxide layer during corrosion process will affect the corrosion resistance of zirconium alloys,the improvement of corrosion resistance of the N18 alloy by obtaining the microstructure with nano-size and uniform distribution of SPPs,and by decreasing the content of Sn and maintaining the content of Cr is discussed.