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...     

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。     

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.     

An oxidation study of Ni-Cr-Fe alloys in carbon dioxide at 800 to 1000° C

Nickel-based alloys containing chromium and iron are alternative high-temperature materials to austenitic stainless steels at temperatures in excess of 800° C in gas-cooled reactor systems. In particular, some of the commercial superalloys may find application as cladding for ceramic fuel elements. These complex alloys contain a number of minor constituents that appear to play an ill-defined but essential role in regulating mechanical properties and oxidation resistance.This paper describes a structural evaluation of thin, oxide films formed on two such alloys, Hastelloy X and Inconel 600, under representative reactor conditions, with emphasis on the application of electron and X-ray diffraction techniques for the identification of multicomponent, surface oxide phases. The difficulties and limitations of these techniques in relation to the complexity of the oxide under examination are apparent, but nevertheless the large amount of useful information obtained has enabled some understanding of the oxidation mechanism. As in the case of highly alloyed, chromiumnickel steels, the rate-determining step in the growth of surface oxide on both alloys appears to be cation diffusion through an initially formed, chromic oxide phase, resulting in the subsequent development of an outer spinel oxide of variable composition.     

Observation of crack microstructure in oxides and its correlation to oxidation and hydrogen-uptake by 3D FIB Tomography – case of Zr-ZrO2 in reactor

Abstract

Zirconium based alloys are used as fuel claddings in Light Water Reactors due to their good resistance to degradation and low neutron absorption cross section. However, life limiting processes occur during the service of the cladding such as oxidation and hydrogen-uptake. During the oxidation of the material, hydrogen enters the metal and it precipitates as brittle hydrides. In this study the 3D microstructure of a high burn-up and a low-burnup LK3/L Zircaloy-2 cladding is characterized and compared using FIB Tomography. 3D reconstruction of the oxides of the claddings shows that the crack volume fraction increases with the number of cycles in the reactor, reducing its protectiveness against further corrosion and H-uptake. The visualization of the metal-oxide interface revealed that the oxidation of the hydrides in the metal could induce crack formation in the oxide and therefore it could be one of the causes of the increasing oxidation and H-uptake in this material.     

Application of neutron radiography to study material processes during hypothetical severe accidents in nuclear reactors

Neutron radiography was applied for investigations of nuclear fuel cladding and control rod behaviour during steam oxidation at temperatures between 1123 and 1673?K under severe nuclear accident conditions. This article gives an overview of these investigations. At KIT, loss of coolant and severe nuclear accidents were experimentally simulated. Post-test examinations of damaged control rods were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. It could be seen clearly from these data that local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube. The uptake of hydrogen during steam oxidation and its diffusion in Zircaloy-4 was investigated in ex situ and in situ radiography experiments at temperatures of 1123 to 1673?K. The kinetics of hydrogen uptake and diffusion was determined. The oxide layer morphology strongly influences the hydrogen concentration in steam oxidized zirconium alloys. Differences of nearly one order of magnitude were found in samples withdrawn from large scale QUENCH experiments. The hydrogen diffusion coefficients were determined for various temperatures. Whereas the diffusion coefficients at 1123 and 1173?K agree well with values expected from literature values for pure Zr, at higher temperatures a faster diffusion was found. The determined activation energy of the hydrogen diffusion is about 10?% higher than published values in the literature.     

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。     

耐事故燃料包壳用FeCrAl不锈钢的研究进展

3.11日本福岛核事故后,国内外围绕提高核燃料元件的事故包容能力和固有安全性课题开展了大量的耐事故燃料(ATF)的研发工作,其中性能更先进的包壳材料是ATF研究的前沿和难点。Fe Cr Al合金优良的高温性能结合管材制备工艺的技术成熟度和经济性,促使该合金包壳成为近期ATF包壳研发的首选目标。简介了国外在ATF包壳候选材料的筛选和Fe Cr Al不锈钢上的研究进展,综述了ATF包壳用Fe Cr Al不锈钢高温蒸汽氧化性能、力学性能、中子辐照性能和应力腐蚀性能等方面的研究现状,指出了Fe Cr Al包壳研制和工程应用等方面需突破的关键技术和研究方向,其中成分优化控制及制备工艺、中子辐照性能和应力腐蚀性能等工程应用的评价是未来研究的重点和难点。     

Wavelength dependent neutron transmission and radiography investigations of the high temperature behaviour of materials applied in nuclear fuel and control rod claddings

Neutron radiography was used for the investigation of the nuclear fuel and control rod cladding behaviour during steam oxidation under severe nuclear accident conditions. In order to verify the hypothesis that the unexpectedly high neutron cross-section found after oxidation of Zircaloy-4 in wet air containing 10% steam is caused by a strong hydrogen uptake, the wavelength dependence of the total macroscopic neutron cross-section of the specimens was measured. The characteristic dependence for hydrogen was not found, which is a proof that hydrogen is not absorbed significantly. The data agree mostly with the behaviour expected for β-Zr.Examinations of control rod simulators annealed until the failure in single-rod tests were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. This procedure clearly showed that the local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube is the reason for the failure.     

氢化锆在高温水蒸气中的氧化行为

通过恒温恒压氧化实验研究了氢化锆在300~700℃高温水蒸气中的氧化行为。结果表明,氢化锆的质量增重随着氧化温度的升高而增大。在氧化过程进行15h以后,OH-/O在氧化膜中的内扩散成为氧化反应的控速步骤。水蒸气中的H抑制了氢化锆在高温氧化条件下的氢损失。氧化膜主要由单斜相M-ZrO2组成,氧化膜的最外层由四方相T-ZrO2和立方相C-ZrO2组成。     

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。     

Fe-Ni基合金设计中前过渡族元素对结构与性能的影响EI北大核心CSCD

铁镍(Fe-Ni)基合金具有优异的高温强度、良好的强韧匹配以及耐腐蚀、抗氧化等特性,在核能、航空、航天领域应用广泛。然而,随着现代工业的飞速发展,材料的服役环境愈发苛刻,传统Fe-Ni基合金无法满足未来重大工程的应用需求。添加合金化元素是优化Fe-Ni基合金使役性能的重要手段,已成为当下的研究热点之一。本文综述了前过渡族元素添加对Fe-Ni合金相结构及性能的影响。基于相图、相变的热力学基础理论,重点讨论前过渡族元素(RE,Ti,V,Cr,Zr,Nb,Mo,Hf,Ta,W等)添加对合金物相结构的影响,论述Fe-Ni基合金成分-结构-服役性能的关系,指出当前元素改性Fe-Ni基合金的研究工作偏向于多元合金协同作用于Fe-Ni基合金,缺乏添加单一元素的研究与相应的热、动力学的数据支撑。提出在明确单一元素添加对Fe-Ni基合金结构及性能作用规律的基础上,探索不同元素的作用机理,建立相应的热、动力学数据库,优化合金在氧化过程中的模拟与计算的解决方法。     

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。     

等离子熔覆法制备AlxCoCrFeNi高熵合金微观组织与性能研究

利用等离子熔覆法在Q235基体上制备了Al_xCoCrFeNi(x=1、1.5,x为摩尔分数)高熵合金,对熔覆层的化学成分、相结构、微观组织和显微硬度进行了研究。结果表明:熔覆态高熵合金具有简单的固溶体结构,微观组织为树枝晶,Al含量从x=1增加到x=1.5时,物相组成由FCC+BCC两相转变为单一的BCC相;当x=1.5时,枝晶间有纳米级颗粒析出;Al_(1.5)CoCrFeNi熔覆层与基体呈现良好的冶金结合,界面附近的热影响区由于珠光体脱碳分解而形成了约为80μm宽的铁素体带;随着Al含量的增加,熔覆层的显微硬度从x=1时的478HV增加到x=1.5时的530HV。     

Atomic-scale interface structure of a Cr-coated Zircaloy-4 material

Highly adherent, thin Cr coatings on Zr-based nuclear fuel claddings can be potentially used for the development of accident-tolerant fuels in light water reactors. To guarantee the successful implementation of Cr-coated Zr alloys as cladding tubes in nuclear power plants, the adhesive strength of the Cr coatings must be assessed. The interface between Cr and Zr was characterized via high-resolution transmission electron microscopy. We observed the formation of nanometer-thick Zr(Fe, Cr)₂ poly-type, structured Laves phases at the interfacial region that display both C14 and C15 lattice symmetries. Although the crystallinity was preserved throughout the interfacial region, different atomic configurations were observed for all the interfaces studied. In most cases, coherent or semicoherent crystallographic relationships were observed, ensuring the adhesive strength of the coating.     

Selection of fuel cladding material for nuclear fission reactors

The growing understanding of the link between carbon emissions and global warming has been promoting a discussion on the environmental and safety viability of nuclear power generation. Current open fuel cycle reactors, however, result in low energy efficiency and produce large volumes of nuclear waste. More advanced nuclear reactors, which are currently under investigation, are expected to allow more efficient and safer use of nuclear energy. In all these cases, the fuel cladding is the most important safety barrier in fission nuclear reactors, as it restrains most of the radioactive fission products within its volume. The selection of fuel cladding material is based on many design constraints, such as neutron absorption cross section, service temperature, mechanical strength, toughness, neutron radiation resistance, thermal expansion, thermal conductivity, and chemical compatibility. The present paper reviews the selection of nuclear fuel cladding materials since the early reactors, illustrating some of the main failure modes and briefly discussing the challenges facing the development of fuel cladding materials for generation IV reactors.     

Modeling of structural failure of Zircaloy claddings induced by multiple hydride cracks

Zirconium alloys have been serving as primary structural materials for nuclear fuel claddings. Structural failure analysis under extreme conditions is critical to the assessment of the performance and safety of nuclear fuel claddings. This work focuses on simulating structural failure of Zircaloy tubes with multiple hydride defects through modeling explicit crack propagation in ductile media. First, we developed an integrated cladding failure model by taking into account both crack initiation induced by hydride/matrix interface separation and ligament tearing-off between activated hydride cracks. Second, to accommodate the initiation, propagation, and coalescence of multiple cracks in finite plastic media we incorporated this structural failure model into a coupled continuous/discontinuous Galerkin (DG) based finite element code, a traditionally preferred implicit numerical framework. Third, to improve the adaptive placement of DG interface elements for crack propagation and to identify potential coalescence of cracks due to the interaction between adjacent hydride cracks, we defined a special failure index for the assessment of potential failure zones using both true plastic strain developed and predicted failure strain based on the Johnson–Cook material failure criterion. Finally, by calibrating the proposed material failure model using a cluster of Zircaloy material experimental tests, we successfully simulated a complete failure process of a fuel cladding tube with multiple hydride cracks.     

Research status of laser cladding on titanium and its alloys: A review

Titanium and its alloys are restricted in the field of application. Laser cladding, as a kind of surface modification technique, is employed to fabricate coatings with improved wear resistance, high temperature oxidation resistance and good biocompatibility on titanium and its alloys. Apart from laser cladding process parameters, material selection is vital to obtain the improved properties mentioned above. In this paper, recent developments of different material system are summarized and the developments in laser cladding for functional coatings with high wear resistance, good corrosion and oxidation resistance, and better medical biocompatibility are reviewed. Besides, the existing problems and the corresponding solutions are discussed. Finally, the trend of development in the future is forecasted.     

Simulation of the coupling behaviors of particle and matrix irradiation swelling and cladding irradiation growth of plate-type dispersion nuclear fuel elements

Within plate-type dispersion nuclear fuel elements, besides irradiation swelling of fuel particles induced by nuclear fissions, the metal matrix and the cladding are attacked continuously by the fast neutrons released from the fuel particles. As a consequence, the matrix undergoes a bit irradiation swelling and the cladding takes on irradiation growth, which both might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, the three-dimensional large-strain constitutive relations for the fuel particles, the metal matrix and cladding are developed; based on them, the method of virtual temperature increase proposed by Ding et al. (2008) is further developed to model the irradiation swelling; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-temperature-loadstep is proposed to simulate the coupling features of the irradiation swellings of both the metal matrix and the fuel particles together with the irradiation growth of the cladding. In order to clarify the critical factors that affect their mechanical performances and carry out optimal design, with the aid of the research thoughts of particle-reinforced composites, numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate the effects of irradiation swelling of the matrix and irradiation growth of the cladding as that: (1) they might weaken the in-pile mechanical performances at the matrix to some extent; and (2) the former increases interfacial stresses between the fuel meat and the cladding, while the latter relatively relieve those interfacial stresses; and the interfacial mechanical strength might be improved by getting suitable irradiation growth mode of the cladding.