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.     

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.     

Effects of helium on the mechanical properties and microstructure of niobium

Mechanical property specimens of niobium (Cb) were doped with helium by the tritium trick to concentrations as high as 500 appm. The tritium decays by the reaction³H →³He +β ⁻ at a rate that produces about 7 appm per day in the host microstructure. Tensile properties were measured from room temperature to 800°C, and creep properties from 700 to 1000°C at stresses from 45 to 75 MPa. Transmission electron microscopy was used to study the microstructure of the helium doped specimens, and the observations were correlated with the mechanical property results. The results of this investigation showed that niobium has a high tolerance to helium trapped in the microstructure. The tensile and creep strengths of niobium increased as helium concentration increased. The ductility decreased significantly as the helium concentration increased, but niobium retained substantial ductility even at a high helium concentration of 500 appm. This amount of helium would be generated by (n,α) reactions in the microstructure of a niobium first wall after a 20 y exposure in a D-T fusion reactor. Thus, niobium and niobium alloys are potential candidates for high temperature structural materials in D-T fusion reactors.     

Corrosion of inconel 617 in HTGR grade helium

Corrosion of the primary circuit materials is one of the serious problems in High Temperature Gas-cooled Reactors (HTGR). In the present work, the effect of gaseous impurities in the helium coolant on the corrosion behavior of Inconel 617 has been studied at 1000 °C and atmospheric pressure for 1000 h. The helium gases used contained several impurities; H₂, H₂O, CH₄, CO, and CO₂. The corrosion behavior of alloying elements (Cr, Mo, Si, C) was strongly affected by the impurity concentration in helium and the extent of corrosion could be explained on the basis of the oxygen and carbon potentials in the gases used. Ryoji Watanabe, formerly Head of Nuclear Materials Division.     

面向等离子体材料用先进钨复合材料的改性研究进展与趋势

核聚变能是解决未来社会能源危机的有效途径之一，但面向等离子体材料在聚变堆体中面临着来自等离子体严重的辐照和热冲击损伤。纯钨因其高热导率、良好的高温强度、低溅射和低蒸气压而被认为是最有希望的面向等离子体候选材料。纯钨在聚变堆工况条件下具有严重的脆性风险，因而对面向等离子体材料用先进钨材料的改性成为近年来的研究热点。钨基复合材料的改性方法主要包括合金化、第二相强化、纤维增韧和复合强化。本文综述了近年来国内外针对核聚变反应堆面向等离子体材料用钨基复合材料的改性方法及其性能，分析了钨基复合材料的改性机制，并展望了面向等离子体材料用钨基复合材料的发展方向。     

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。      

Al2O3 Regions/Grains in ODS Steel PM2000 Irradiated With Fe Ions at 700 °C

Metallurgical and Materials Transactions A - Oxide dispersion strengthened (ODS) steel is one of the candidate structural materials for Generation-IV nuclear reactors. The microstructure of...     

抗辐照合金的发展与纳米结构ODS钢的抗辐照性能

简要介绍了抗辐照合金的发展,合金微观结构对合金抗辐照性能的影响,先进堆核心部件结构材料最佳的备选材料纳米结构氧化物弥散强化钢的特征性微观结构及其抗辐照性能.     

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.     

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.     

Irradiation Behavior in High Entropy Alloys

As an increasing demand of advanced nuclear fission reactors and fusion facilities,the key requirements for the materials used in advanced nuclear systems should encompass superior high temperature property,good behavior in corrosive environment,and high irradiation resistance,etc.Recently,it was found that some selected high entropy alloys(HEAs) possess excellent mechanical properties at high temperature,high corrosion resistance,and no grain coarsening and self-healing ability under irradiation,especially,the exceptional structural stability and lower irradiation-induced volume swelling,compared with other conventional materials.Thus,HEAs have been considered as the potential nuclear materials used for future fission or fusion reactors,which are designed to operate at higher temperatures and higher radiation doses up to several hundreds of displacement per atom(dpa).An insight into the irradiation behavior of HEAs was given,including fundamental researches to investigate the irradiation-induced phase crystal structure change and volume swelling in HEAs.In summary,a brief overview of the irradiation behavior in HEAs was made and the irradiation-induced structural change in HEAs may be relatively insensitive because of their special structures.     

Failure mechanisms in SiC-fiber reinforced 6061 aluminum alloy composites under monotonic and cyclic loading

Micromechanisms influencing crack propagation in a unidirectional SiC-fiber (SCS-8) continuously reinforced Al-Mg-Si 6061 alloy metal-matrix composite (SiC_f/Al-6061) during monotonie and cyclic loading are examined at room temperature, both for the longitudinal (0 deg or L-T) and transverse (90 deg or T-L) orientations. It is found that the composite is insensitive to the presence of notches in the L-T orientation under pure tension loading due to the weak fiber/matrix interface; notched failure strengths are ∼1500 MPa compared to 124 MPa for unreinforced 6061. However, behavior is strongly dependent on loading configuration, specimen geometry, and orientation. Specifically, properties in SiC_f/Al in the T-L orientation are inferior to unreinforced 6061, although the composite does exhibit increasing crack-growth resistance with crack extension (resistance-curve behavior) under monotonie loading; peak toughnesses of ∼16 MPa√m are achieved due to crack bridging by the continuous metal phase between fibers and residual plastic deformation in the crack wake. In contrast, such bridging is minimal under cyclic loading, as the ductile phase fails subcritically by fatigue such that the transverse fatigue crack-growth resistance is superior in the unreinforced alloy, particularly at high stress-intensity levels. Conversely, fatigue cracks are bridged by unbroken SiC fibers in the L-T orientation and exhibit marked crack deflection and branching; the fatigue crack-growth resistance in this orientation is clearly superior in the composite.     

高强度耐腐蚀ODS?FeCrAl合金微观结构、力学性能研究进展

氧化物弥散强化（Oxide dispersion strengthened，ODS）FeCrAl合金由于加入一定量的Al元素，使合金表面可形成一层薄而致密的Al₂O₃保护膜，使得合金即便在1400 ℃的水蒸汽下也不会因为腐蚀导致失效。同时，大量超细氧化物粒子的弥散强化作用使其具备优异的高温强度。这种兼具高温强度和耐腐蚀的特性使得ODS?FeCrAl合金成为非常有前景的事故容错燃料（Accident tolerant fuel , ATF）包壳候选材料，也是快堆等其他工作于高温强腐蚀环境的先进反应堆包壳的重要候选材料。Al元素的引入会使ODS铁基合金中弥散粒子的种类发生变化，进而影响其显微组织和力学性能。针对ODS?FeCrAl合金中引入Al元素所导致的显微组织变化及其对蠕变性能的影响，总结了国内外相关研究进展，旨在为适用于先进反应堆的ODS?FeCrAl合金的发展提供参考。      

High temperature creep-fatigue design

Generation IV fission and future fusion reactors envisage development of more efficient high temperature concepts where materials performances are key to their success. This paper examines different types of high temperature creep-fatigue interactions and their implications on design rules for the structural materials retained in both programmes. More precisely, the paper examines current status of design rules for the stainless steel type 316L(N), the conventional Modified 9Cr-1Mo martensitic steel and the low activation Eurofer steel. Results obtained from extensive high temperature creep, fatigue and creep-fatigue tests performed on these materials and their welded joints are presented. These include sequential creep-fatigue and relaxation creep-fatigue tests with hold times in tension, in compression or in both. Effects of larger plastic deformations on fatigue properties are studied through cyclic creep tests or fatigue tests with extended hold time in creep. In most cases, mechanical test results are accompanied with microstructural and fractographic observations. In the case of martensitic steels, the effect of oxidation is examined by performing creep-fatigue tests on identical specimens in vacuum. Results obtained are analyzed and their implications on design allowables and creep-fatigue interaction diagrams are presented. While reasonable confidence is found in predicting creep-fatigue damage through existing code procedures for austenitic stainless steels, effects of cyclic softening and coarsening of microstructure of martensitic steels throughout the fatigue life on materials properties need to be taken into account for more precise damage calculations. In the long-term, development of ferritic/martensitic steels with stable microstructure, such as ODS steels, is proposed.     

Analysis of interfacial shear strength of SiC fiber reinforced 7075 aluminum composite by pushout microindentation

The interfacial shear strength of continuous silicon carbide fiber reinforced 7075 aluminum matrix composite (SiC_f/7075Al) has been investigated in this research by pushout microindentation. The SiC_f/7075Al composite specimens were processed by diffusion bonding alternate layers of SiC fibers and 7075Al alloy plates. From the measured stress-displacement curves of indentation tests, the interfacial shear strengths of the composite specimens were obtained, and the stress-displacement curves were basically divided into two regions: (1) elastic deformation and (2) interface decohesion and fiber sliding. With increasing aging time, the interfacial shear strength of the composite increased to 167 MPa for T6-treated specimens, and the variation of the interfacial shear strength well followed that of the ultimate tensile strength of 7075Al matrix alloy. With decreasing specimen thickness, the interfacial shear strength of the composite and the amplitude of stress fluctuation slightly decreased because of the stress relaxation effect near specimen surfaces. Under higher indentation velocities, both the interfacial shear strength and the amplitude of stress fluctuation became smaller.     

The effects of reinforcement additions and heat treatment on the evolution of the poisson ratio during straining of discontinuously reinforced aluminum alloys

The effects of reinforcement additions and heat treatment on the evolution of the Poisson ratio were determined for a 7xxx aluminum alloy reinforced with 15 vol pct SiC_p, a 2xxx alloy with 20 pct SiC_p, and a 2014 alloy with 15 pct A1₂O₃. The Poisson ratio of the monolithic alloy was 0.31 to 0.32 in the elastic regime. At the onset of the plastic regime, the Poisson ratio of the monolithic materials rose rapidly to about 0.45 and then gradually increased to 0.47 by 3.5 pct strain. For discontinuously reinforced aluminum (DRA) materials, the Poisson ratio in the elastic regime was considerably lower than that exhibited by the matrix alloy, while the magnitude of the difference was dependent upon the type, volume fraction, and elastic properties of the reinforcement. In addition, th evolution of the Poisson ratio for DRA material depends upon heat treatment and level of strain due to damage evolution(e.g., SiC_p cracking, matrix failure,etc.) which accompanies straining in these materials. Both the magnitude and extent of change in the Poisson ratio with increasing strain in the composite is rationalized by the accumulation of damage which accompanies increasing strain.     

Superplasticity in a high strength powder aluminum

The superplastic properties of a rapidly solidified, high strength P/M Al alloy and the same alloy reinforced with SiC particulates (SiC _p ) have been studied. To prepare superplastic test materials, a matrix alloy powder of composition 7.2Zn-2.4Mg-2Cu-0.2Zr-0.12Cr-0.2Co (Kaiser PM-64) and the powder mixed with 10 to 20 vol pct SiC _p (~5 μm diameter) were thermomechanically processed to very fine equiaxed grain structures of ~6 μm and ~8 μm, respectively. Superplasticity in these materials was evaluated by characterizing (1) high temperature stability, (2) dynamic grain growth, (3) strain rate sensitivity, (4) flow stress behavior, (5) cavitation and cavitation control, and (6) total superplastic strain. It was observed that the PM-64 alloy could achieve a total elongation of over 800 pct, while the SiC_p reinforced alloy could attain an elongation greater than 500 pct before failure. Also, it was shown that with the use of hydrostatic pressure during superplastic flow, cavitation could be controlled. Observations were made of the effect SiC _p reinforcement particles had on the superplastic flow stress behavior. Interpretations are proposed to explain the role of particulates during superplastic straining.     

Relationships between phase stability and void swelling in Fe-Cr-Ni alloys during irradiation

Austenitic Fe-Cr-Ni steels are potential candidate alloys for structural materials in both fast breeder and magnetic fusion reactors. However, void swelling and phase instability during irradiation have been major problems limiting the properties and useful lifetimes of these materials and thus have been the subject of intensive investigations. Cavity nucleation in steels subjected to displacement damage is strongly influenced by the interactions between helium atoms and precipitates which are formed during irradiation. Several important mechanisms regarding gas atom-precipitate interactions and principles for the design of radiation-resistant alloys are critically examined. Central concepts derived from theory, including the critical radius and critical number of gas atoms, cavity-precipitate interactions, and relative sink strengths for point defects, are applied to the interpretation of experimental data. This paper is based on a presentation made in the “G. Marshall Pound Memorial Symposium on the Kinetics of Phase Transformations” presented as part of the 1990 fall meeting of TMS, October 8–12, 1990, in Detroit, MI, under the auspices of the ASM/MSD Phase Transformations Committee.     

Contributions from research on irradiated ferritic/martensitic steels to materials science and engineering

Ferritic and martensitic steels are finding increased application for structural components in several reactor systems. Low-alloy steels have long been used for pressure vessels in light water fission reactors. Martensitic stainless steels are finding increasing usage in liquid metal fast breeder reactors and are being considered for fusion reactor applications when such systems become commercially viable. Recent efforts have evaluated the applicability of oxide dispersion-strengthened ferritic steels. Experiments on the effect of irradiation on these steels provide several examples where contributions are being made to materials science and engineering. Examples are given demonstrating improvements in basic understanding, small specimen test procedure development, and alloy development. 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.