材料辐照损伤中的点缺陷团簇与一维迁移现象

材料辐照损伤是核反应堆材料、尤其是核聚变堆材料所面临的重要问题。高能粒子（中子、离子、电子）辐照在材料中会产生大量的点缺陷,即自间隙原子和空位。这些点缺陷聚集在一起会形成自间隙原子团簇和空位团簇,从而对材料结构和性能的演化产生重要影响。空位团簇包括有空洞、层错四面体、空位型位错环,而自间隙原子团簇则只有自间隙型位错环。本文介绍了两种点缺陷团簇的性质、及其对于以材料辐照肿胀为主要内容的材料辐照损伤性能的影响。作为空位团簇,比较详细介绍了具有本课题组特色的空位型位错环的研究,同时分析了合金元素和氢同位素对空位型位错环的影响。在铁试样中形成的这种空位型位错环尺寸可达100 nm左右,该空位型位错环具有两种柏氏矢量, b =<100> 和 b =1/2<111>,前者的数密度比后者高一个数量级。对于自间隙原子团簇,则重点介绍了与其相关的一维迁移现象及其研究动态,该一维迁移性能有可能是影响高熵合金辐照性能的重要因素。      

Thermal and Irradiation Creep Behavior of a Titanium Aluminide in Advanced Nuclear Plant Environments

Titanium aluminides are well-accepted elevated temperature materials. In conventional applications, their poor oxidation resistance limits the maximum operating temperature. Advanced reactors operate in nonoxidizing environments. This could enlarge the applicability of these materials to higher temperatures. The behavior of a cast gamma-alpha-2 TiAl was investigated under thermal and irradiation conditions. Irradiation creep was studied in beam using helium implantation. Dog-bone samples of dimensions 10 × 2 × 0.2 mm³ were investigated in a temperature range of 300 °C to 500 °C under irradiation, and significant creep strains were detected. At temperatures above 500 °C, thermal creep becomes the predominant mechanism. Thermal creep was investigated at temperatures up to 900 °C without irradiation with samples of the same geometry. The results are compared with other materials considered for advanced fission applications. These are a ferritic oxide-dispersion-strengthened material (PM2000) and the nickel-base superalloy IN617. A better thermal creep behavior than IN617 was found in the entire temperature range. Up to 900 °C, the expected 10⁴ hour stress rupture properties exceeded even those of the ODS alloy. The irradiation creep performance of the titanium aluminide was comparable with the ODS steels. For IN617, no irradiation creep experiments were performed due to the expected low irradiation resistance (swelling, helium embrittlement) of nickel-base alloys.     

Materials requirements for fusion reactors

Once the physics of fusion devices is understood, which is expected to be achieved in the early 1980’s, one or more experimental power reactors (EPR) are planned which will pro-duce net electrical power. The structural material for the device will probably be a modi-fication of an austenitic stainless steel. Unlike fission reactors, whose pressure bound-aries are subjected to no or only light irradiation, the pressure boundary of a fusion reac-tor is subjected to high atomic displacement-damage and high production rates of trans-mutation products,e.g., helium and hydrogen. Hence, the design data base must include irradiated materials. Sincein situ testing to obtain tensile, fatigue, creep, crack-growth, stress-rupture, and swelling data is currently impossible for fusion reactor conditions, a program of service-temperature irradiations in fission reactors followed by postirradi-ation testing, simulation of fusion conditions, and low-fluence 14 MeV neutron-irradiation tests are planned. For the Demonstration Reactor (DEMO) expected to be built within ten years after the EPR, higher heat fluxes may require the use of refractory metals, at least for the first 20 cm. A partial data base may be provided by high-flux 14 MeV neutron sources being planned. Many materials other than those for structural components will be required in the EPR and DEMO. These include superconducting magnets, insulators, neutron reflectors and shields, and breeding materials. The rest of the device should utilize conventional materials except that portion involved in tritium confinement and re-covery. This paper is based on a presentation made at a symposium on “Materials Re-quirements for Unconventional Energy Systems” held at the Niagara Falls meeting of The Metallurgical Society of AIME, September 22, 1976, under the sponsorship of Non-Ferrous Metals and Ferrous Metals Committees.     

A Brief Review of High Entropy Alloys and Serration Behavior and Flow Units

Multicomponent alloys with high entropy of mixing,e.g.,high entropy alloys (HEAs)and/or multiprin-cipal-element alloys (MEAs),are attracting increasing attentions,because the materials with novel properties are being developed,based on the design strategy of the equiatomic ratio,multicomponent,and high entropy of mixing in their liquid or random solution state.Recently,HEAs with the ultrahigh strength and fracture toughness,excel-lent magnetic properties,high fatigue,wear and corrosion resistance,great phase stability/high resistance to heat-softening behavior,sluggish diffusion effects,and potential superconductivity,etc.,were developed.The HEAs can even have very high irradiation resistance and may have some self-healing effects,and can potentially be used as the first wall and nuclear fuel cladding materials.Serration behaviors and flow units are powerful methods to understand the plastic deformation or fracture of materials.The methods have been successfully used to study the plasticity of amorphous alloys (also bulk metallic glasses,BMGs).The flow units are proposed as:free volumes,shear transi-tion zones (STZs),tension-transition zones (TTZs),liquid-like regions,soft regions or soft spots,etc.The flow units in the crystalline alloys are usually dislocations,which may interact with the solute atoms,interstitial types,or sub-stitution types.Moreover,the flow units often change with the testing temperatures and loading strain rates,e.g., at the low temperature and high strain rate,plastic deformation will be carried out by the flow unit of twinning,and at high temperatures,the grain boundary will be the weak area,and play as the flow unit.The serration shapes are related to the types of flow units,and the serration behavior can be analyzed using the power law and modified power law.     

高剂量氦离子辐照对新型中子增殖铍钨合金表面结构的影响

为了确保未来核聚变反应堆的氘氚自持燃烧必需采用中子增殖材料来得到合适的氚增值比。金属铍被认为是最有前途的核聚变反应堆固态中子倍增材料,但其熔点低,高温抗辐照肿胀性能差,因此需要寻找和研发具有更高熔点和更耐辐照肿胀的新型中子增殖材料以满足更先进的聚变堆要求。本研究尝试提出并制备了一种更高熔点的铍钨合金（Be₁₂W）,通过X射线和扫描电子显微镜对它的相组成和表面结构进行分析。对新型铍钨合金进行高剂量的氦离子辐照,发现合金表面一次起泡的平均尺寸约为0.8 μm,面密度约为2.4×107 cm?2,而二次起泡的平均尺寸约为80 nm,面密度约为1.28×108 cm?2。分析氦辐照引起的表面起泡及其机制,并与纯铍和铍钛合金表面起泡的情况进行了对比。      

Ferritic/martensitic steels for advanced nuclear reactors

Design concepts for the next generation of nuclear power reactors include water-cooled, gas-cooled, and liquid-metal-cooled reactors. Reactor conditions for several designs offer challenges for engineers and designers concerning which structural and cladding materials to use. Depending on operating conditions, some designs favor elevated-temperature ferritic/martensitic steels for in-core and out-of core applications. Such steels have been investigated in previous work on international fast reactor and fusion reactor research programs. Steels from these fission and fusion programs will provide reference materials for future fission applications. In addition, new elevated-temperature steels have been developed in recent years for conventional power systems that also need to be considered.     

高熵合金耐腐蚀性能研究进展

文章介绍了高熵合金块体材料和高熵合金涂层在常温环境中、高温条件下和一些特殊介质中的腐蚀行为,论述了合金元素、热处理、环境因素及制备工艺对高熵合金耐蚀性的影响,并简要分析了高熵合金耐蚀性研究面临的问题.     

Structural materials for fusion reactors

Fusion Reactors will require specially engineered structural materials, which will simultaneously satisfy the harsh conditions such as high thermo mechanical stresses, high heat loads and severe radiation damage without compromising on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics of structural materials. There exists considerable uncertainty in the nuclear data at such energies because there aren’t any strong enough sources for such neutrons except fusion reactors themselves! We thus encounter a problem of iterative nature in which we must try several experiments with the available materials in the near term. The development of such structural materials is thus going to require the experimental data of the kind that may be generated on reactors like ITER, high-performance modeling and a penetrating metallurgical insight to overcome technological challenges in terms of achieving required properties such as low activation by controlling the impurities, good thermo-mechanical properties by microstructure engineering, good chemical compatibility and high radiation resistance. These materials need to withstand a neutron wall load of the order of 2–3 MW/m², which can lead up to 30 dpa of radiation damage and 300 appm helium production per full power year in DEMO like reactors. Such conditions lead to unprecedented events related to the failure of materials due to irradiation creep, Ductile-Brittle Transition Temperature (DBTT) shift and helium embrittlement. The development of fusion materials program is oriented towards fulfilling the requirements of Test Blanket Modules, various prototype activities of SST-2 and DEMO reactor. The materials identified for first wall and blanket modules for Indian DEMO are LAFMS and ODS steels. The development program plan for these materials include (i) Manufacturing of LAFMS steel through VIM/VAR methods by controlling the impurities such as S, P and Si. (ii) ODS steel development with nano-size Y₂O₃ dispersoids in ferritic martensitic matrix by powder metallurgy route. The advanced structural materials like SiC_f /SiC composites and SiC_f /n-SiC are planned under National Fusion Program projects for indigenous development. An overview of the planned program in this direction will be presented.     

Irradiation Resistance of Multicomponent Alloys

High-entropy alloys (HEAs) are characterized not only by high values of entropy but also by high atomic-level stresses originating from mixing of elements with different atomic sizes. Particle irradiation on solids produces atomic displacements and thermal spikes. The high atomic-level stresses in HEAs facilitate amorphization upon particle irradiation, followed by local melting and re-crystallization due to thermal spikes. We speculate that this process will leave much less defects in HEAs than in conventional alloys. For this reason, they may be excellent candidates as new nuclear materials. We discuss initial results of computer simulation on model binary alloys and an electron microscopy study on Zr-Hf-Nb alloys, which demonstrate extremely high irradiation resistance of these alloys against electron damage to support this speculation.     

Creep and fatigue issues for structural materials in demonstration fusion energy systems

The structural materials proposed for use in future fusion energy systems must perform reliably in an environment consisting of intense neutron irradiation, high temperatures, and cyclic stress. Therefore, thermal creep and creep-fatigue (in addition to irradiation creep) are anticipated to be important issues for the engineering design of structural materials for fusion reactors. The key materials systems under consideration for structures of fusion reactors include 8–9%Cr ferritic/martensitic steels, oxide dispersion strengthened ferritic steels, vanadium alloys and SiC fiber-reinforced SiC matrix ceramic composites. The current elevated temperature creep-fatigue design rules based on the American Society of Mechanical Engineers (ASME) code are discussed, along with a brief review of creep-fatigue interaction mechanisms. Refinements to current international design codes to include radiation-induced phenomena such as reduction in uniform elongation have been performed in association with the engineering design of the ITER fusion energy device currently under construction in France. Several other creep-fatigue issues of potential importance for fusion energy applications are discussed.     

难熔高熵合金:制备方法与性能综述

从加工方法、微观结构以及各类性能三方面介绍了难熔高熵合金(Refractory high-entropy alloys,RHEAs),最后对难熔高熵合金的发展和未来进行了展望。以MoNbTaVW为代表的难熔高熵合金在高温下表现出优于传统镍基高温合金的压缩屈服强度,且屈服强度随温度的变化更加缓慢,高温力学性能优异;以MoNbTaVW、MoNbTaTiZr、HfNbTiZr等为代表的难熔高熵合金,与商用高温合金、难熔金属、难熔合金以及工具钢相比,展现出更优的耐磨性能。以W₃₈Ta₃₆Cr₁₅V₁₁合金为代表的难熔高熵合金在辐照后,除了析出小颗粒第二相外,不存在位错环缺陷结构,抗辐照性能优异。提出了难熔高熵合金未来发展的两大方向:建立高通量的实验和计算方法继续探索更多的难熔高熵合金组成和结构模型;探索多场耦合环境下难熔高熵合金的服役行为。      

粉末冶金在高熵材料中的应用

高熵材料是近年来出现的一种新型材料,具有高强度、高硬度、良好耐腐蚀和优异的高温组织稳定性等性能,在航空航天、高温以及先进核能等领域展现了广阔的应用前景,引起国际材料领域的广泛关注,相关研究也取得了很大进展。粉末冶金作为一种高性能金属基和陶瓷复合材料的先进制备技术,可以获得纳米晶和过饱和固溶体等亚稳材料,同时也可用于传统熔炼法较难制备的具有特殊结构和性能的材料,近些年来,粉末冶金技术在高熵材料制备中得到广泛应用。本文从高熵材料的应用理论出发,针对目前高熵材料粉体制备方法、块体成型以及粉末冶金制备的典型高熵材料三个方面予以综述,着重阐述了高熵材料的力学性能和其变形行为特点,同时展望了高熵材料的未来发展趋势。      

用于超临界水堆燃料包壳的ODS铁素体钢的研究进展

 超临界水堆具有热效率高、系统简化、成本低等优点,成为第四代核反应堆中优先发展的堆型。ODS铁素体钢由于其优异的高温力学性能和良好的抗辐照能力成为超临界水堆包壳最有希望的候选材料。本文旨在回顾ODS铁素体钢制造工艺,包括机械合金化参数的优化,热处理工艺的选择以消除力学性能上的各向异性。根据超临界水堆包壳的服役条件,结合最新的实验数据,对ODS铁素体钢的高温力学性能、在超临界水中的耐腐蚀性以及中子辐照稳定性进行了总结和展望。     

Review of small specimen test techniques for irradiation testing

Small specimen test technology has evolved out of the necessity to develop and monitor materials proposed for or used in nuclear power generation systems. Development of materials for improved cladding and in-core structures for fission reactors and assessment of core materials and pressure vessel steels already under irradiation necessitated the use of specimens which fit into existing irradiation space or which could be extracted from irradiated structures, such as cladding or ducts. Interest in simulating neutron irradiation by light and heavy ion irradiation led to the development of thin foil and wire geometry specimens. Further, interest in developing materials for fusion reactors has added additional constraints on specimen sizes associated with available irradiation volumes in existing and proposed high-energy neutron irradiation facilities. Consequently, a wide array of specimen geometries and test techniques has now been developed. It is the purpose of this paper to review these techniques and examine their status, problems, and potential for future applications. This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25–29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD.     

Review of small specimen test techniques for irradiation testing

Small specimen test technology has evolved out of the necessity to develop and monitor materials proposed for or used in nuclear power generation systems. Development of materials for improved cladding and in-core structures for fission reactors and assessment of core materials and pressure vessel steels already under irradiation necessitated the use of specimens which fit into existing irradiation space or which could be extracted from irradiated structures, such as cladding or ducts. Interest in simulating neutron irradiation by light and heavy ion irradiation led to the development of thin foil and wire geometry specimens. Further, interest in developing materials for fusion reactors has added additional constraints on specimen sizes associated with available irradiation volumes in existing and proposed high-energy neutron irradiation facilities. Consequently, a wide array of specimen geometries and test techniques has now been developed. It is the purpose of this paper to review these techniques and examine their status, problems, and potential for future applications. This paper is based on a presentation made in the symposium “Irradiation-Enhanced Materials Science and Engineering” presented as part of the ASM INTERNATIONAL 75th Anniversary celebration at the 1988 World Materials Congress in Chicago, IL, September 25–29, 1988, under the auspices of the Nuclear Materials Committee of TMS-AIME and ASM-MSD.     

快中子反应堆核心结构材料的辐照损伤

快中子反应堆(快堆)的核心结构材料(如燃料包壳等)在服役过程中将承受长期的高通量的中子辐照、高温和嬗变反应产生的He的作用,引起的合金微观结构的改变,导致材料力学性能的严重恶化.高性能抗辐照材料成为快堆发展的关键前提条件之一.本文介绍快堆中辐照引起的金属材料微观结构的变化.     

Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at <0.3T _M (T _M is melting temperature) and up to 10 dpa (displacement per atom). Ferritic/martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.     

Materials requirements for liquid metal fast breeder reactors

Materials requirements for Liquid Metal Fast Breeder Reactors (LMFBRs) are quite varied with requisite applications ranging from ex-reactor components such as piping, pumps, steam generators and heat exchangers to in-reactor components such as heavy section reactor vessels, core structurais, fuel pin cladding and subassembly flow ducts. Requirements for ex-reactor component materials include: good high temperature ten-sile, creep and fatigue properties; compatibility with high temperature flowing sodium; resistance to wear, stress corrosion cracking, and crack propagation; and good welda-bility. Requirements for in-reactor components include most of those cited above for ex-reactor components as supplemented by the following: resistance to radiation embrittle-ment, swelling and radiation enhanced creep; good neutronics; compatibility with fuel and fission product materials; and resistance to mass transfervia flowing sodium. Extensive programs are currently in place in a number of national laboratories and industrial con-tractors to address the materials requirements for LMFBRs. These programs are fo-cused on meeting the near term requirements of early LMFBRs such as the Fast Flux Test Facility and the Clinch River Breeder Reactor as well as the longer term require-ments of larger near-commercial and fully-commercial reactors.     

Microstructural evolution in Zr3Al-based alloys during various long-time annealing treatments

The Zr₃Al-based intermetallics have been considered as candidate materials for structural components in pressurized heavy water nuclear reactors due to their attractive properties. However, these alloys could not be used for structural application in the reactors because of their poor room-temperature ductility and irradiation-induced amorphization. Our recent studies of ternary addition of niobium in Zr₃Al have shown some promising results. The present article reports the microstructural evolution in these alloys upon long-time annealing treatments. The formation of various phases, temperature regime of their stability, chemical composition, and volume fraction of these phases during prolonged annealing have also been studied. A pseudobinary phase diagram with varying niobium concentration has also been developed. The morphology and distribution of the Zr₃Al phase have been explained on the basis long-range diffusion as the rate-controlling step. This article is based on a presentation made in the symposium entitled “Processing and Properties of Structural Materials,” which occurred during the Fall TMS meeting in Chicago, Illinois, November 9–12, 2003, under the auspices of the Structural Materials Committee.     

RAFM钢应变补偿本构关系及热加工图

低活化铁素体/马氏体(RAFM)钢具有较低的辐照肿胀率和优异的力学性能,被认为是聚变堆首选的结构材料。然而,低活化钢强度高、冷塑性变形抗力大的特点,使其难以通过冷加工或低温加工实现大规模生产。使用MMS-200型热模拟试验机,在变形温度为950～1 200℃、应变速率为0.1～5 s^-1和最大变形量为50%条件下,进行了低活化铁素体/马氏体钢(0.11C-9.4Cr-1.35W-0.22V-0.05Si-0.11Ta-0.50Mn)单道次热压缩试验,研究其热变形行为。基于动态材料模型构建了不同应变量下的低活化钢变形本构方程和热加工图,确定了最优热加工参数,结合金相结果分析了材料变形过程中微观组织演化规律,为低活化钢的热加工成形工艺及组织优化提供理论参考。结果表明,在相同应变速率下,随着变形温度升高,流变应力逐渐降低,在一定变形温度下,流变应力随应变速率增大而增大;温度和应变速率对组织的影响主要取决于变形过程中材料内部发生的动态回复和再结晶等机制的交互作用。使用六阶多项式拟合进行应变补偿建立的低活化钢变形本构方程具有较高的预测精度,平方相关系数为0.972。显微组织...