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Vanadium for Nuclear Systems
Abstract
Vanadium alloys are recognized as attractive candidate materials for the structural component of fusion reactors because of their low activation properties, high thermal stress factors, and radiation resistance. V-4Cr-4Ti has been regarded as the leading candidate composition. Recent Japanese, US, Russian, and Chinese fusion programs largely enhanced the fabrication technology of vanadium alloys. Also, fundamental understandings of the effects of interstitial impurities (C, N, O) on mechanical properties, radiation effects on microstructure and mechanical properties, and compatibility in various environments were advanced recently. The effects of neutron irradiation with transmutant helium on mechanical properties and irradiation/thermal creep performance are among the major remaining issues.
... Vanadium alloys play a key role, as they are contemplated for most advanced DEMO designs using liquid Li as the breeding and cooling material ( Figure 14) due to their good compatibility with this element [88][89][90][91][92]. In addition, these alloys exhibit good resistance to corrosion and irradiation swelling and, in particular, maintain high temperature resistance. ...
... The T retention characteristics of vanadium alloys leave much to be desired, as they possess a H permeability of at least two orders of magnitude more than any other blanket material and can form detrimental hydrides [88,94]. In addition, their high diffusivity and solubility coefficients create a serious problem as the embrittlement of the materials by H is something to be avoided at all costs since it contributes to their degradation [89,91]. ...
... Illustration of a self-cooled Li blanket with structural material V-4Cr-4Ti. Reprinted with permission from Ref.[88]. Credit c Elsevier, 2012. ...
This paper presents the roadmap of the main materials to be used for ITER and DEMO class reactors as well as an overview of the most relevant innovations that have been made in recent years. The main idea in the EUROfusion development program for the FW (first wall) is the use of low-activation materials. Thus far, several candidates have been proposed: RAFM and ODS steels, SiC/SiC ceramic composites and vanadium alloys. In turn, the most relevant diagnostic systems and PFMs (plasma-facing materials) will be described, all accompanied by the corresponding justification for the selection of the materials as well as their main characteristics. Finally, an outlook will be provided on future material development activities to be carried out during the next phase of the conceptual design for DEMO, which is highly dependent on the success of the IFMIF-DONES facility, whose design, operation and objectives are also described in this paper.
... According to the waste disposal and materials recycling criteria, these V-4Cr-4Ti products met the requirement for low activation [16]. Figure 1 shows the comparison of estimated contact dose rates after use in the first wall of a fusion commercial reactor for SS316LN-IG, RAFM steel (F82H), V-4Cr-4Ti (NH2) and pure SiC/SiC [4,19], respectively. ...
... F82H, V-4Cr-4Ti alloy and pure SiC/ SiC may satisfy the full-remote recycling limit. In particular, V-4Cr-4Ti alloy exhibits a significantly reduced dose rate after cooling for 5-10 years [19]. ...
... Design of a self-cooled liquid breeder blanket can simplify the blanket structure significantly. Adoptive liquid metals such as Li-Pb and liquid Li can further improve the capability of heat transfer and chemical stability [19]. However, in a self-cooled blanket design of a magnetic confinement fusion (MCF) system, Magneto hydrodynamic (MHD) pressure drop is a little high and makes the mechanical stresses over the limits of structural materials [31]. ...
Low-activation vanadium alloys, with the reference composition of V–4Cr–4Ti have been considered as one of the most promising candidate materials for structural components such as the blanket in future fusion reactors, thanks to their excellent neutron irradiation resistance, superior high-temperature mechanical properties, and high compatibility with liquid lithium blankets. The self-cooled liquid lithium blanket using structural materials of vanadium alloys is an attractive concept because of the high heat transfer and high tritium breeding capability. After more than 2 decades of research, technological progress has been made in reducing the number of critical issues for application of vanadium alloys to fusion reactors. In this paper, the recent research and development activities of vanadium alloys are summarized, including significant progress achieved on fabrication technology and composition optimization, coating and corrosion, improved understanding of irradiation effects upon microstructure and material properties, retention of hydrogen isotopes, as well as advancements in joining and welding. In particular, the fact that recent products from China, Japan, US and France showed similar properties which meant the fabrication technology has been almost standardized.
... After 850 C and 950 C heat treatment, the aluminized coatings had similar double-layer structure with thickness of the Al 3 (V, Cr, Ti) outer and Al 8 (V, Cr, Ti) 5 inner layer gradually decreased and increased respectively, but Al 3 (V, Cr, Ti) and Al 8 (V, Cr, Ti) 5 appeared in the outer layer at 950 C (Fig. 8c'). For the aluminized coating obtained at 1050 C, the Al concentration remained 43e53 at.% in the outer layer, and 20e35 at.% in the inner layer (Fig. 8d'), corresponding to Al 8 (V, Cr, Ti) 5 and V(Al, Cr, Ti) respectively in the V-Al phase diagram [31], Thus the aluminized coating obtained at 1050 C could be mainly consisted of an outer Al 8 (V, Cr, Ti) 5 layer and an inner V(Al, Cr, Ti) layer (Fig. 8d'). Fig. 9 shows XRD patterns of the above aluminized coatings obtained at 750 C, 850 C, 950 C and 1050 C, respectively. ...
... Accordingly, For 20 mm Al coated specimens, the structure of the aluminized coatings obtained at 750 C, 850 C was Al 3 (V, Cr, Ti)/Al 8 (V, Cr, Ti) 5 . After 950 C heat treatment, the aluminized coating was mainly Al 8 (V, Cr, Ti) 5 , with a few Al 3 (V, Cr, Ti) and that of 1050 C heat treatment was mainly V(Al, Cr, Ti), with a few Al 8 (V, Cr, Ti) 5 and a-Al 2 O 3 scale. ...
... Accordingly, For 20 mm Al coated specimens, the structure of the aluminized coatings obtained at 750 C, 850 C was Al 3 (V, Cr, Ti)/Al 8 (V, Cr, Ti) 5 . After 950 C heat treatment, the aluminized coating was mainly Al 8 (V, Cr, Ti) 5 , with a few Al 3 (V, Cr, Ti) and that of 1050 C heat treatment was mainly V(Al, Cr, Ti), with a few Al 8 (V, Cr, Ti) 5 and a-Al 2 O 3 scale. ...
Preparation of aluminide coatings, AlxVy, is a key step for the formation of V-Al/Al2O3 as tritium permeation barrier (TPB) on V alloys. We firstly attempted to process aluminide coatings on V–5Cr–5Ti by a two-step process of electrochemical Al deposition (ECA) and heat treatment. Al deposition from AlCl3-1-ethyl-3-methyl-imidazolium chloride (AlCl3-EMIC) ionic liquid was performed at room temperature, with current density varying from 8 to 32 mA/cm2. Homogeneous and well adherent Al coatings on V–5Cr–5Ti were obtained by galvanostatic electrodeposition at 20 mA/cm2, with the deposition rates up to 21 μm/h. And then Al-coated specimens were heated in flowing Argon gas, forming aluminide coatings. The aluminide coatings exhibited a double-layer or almost a single-layer structure, and their thickness increased with temperature, time and Al-coating thickness, respectively. Main phase of coatings was Al3(V, Cr, Ti) at initial diffusion stage, and then gradually transformed into Al8(V, Cr, Ti)5, and next to V(Al, Cr, Ti) solid solution, with double-layer structure transforming into single-layer structure. The homogenous layer of Al8(V, Cr, Ti)5 is the appropriate aluminide for further selective oxidation. The optimum aluminized conditions and growth kinetics of Al8(V, Cr, Ti)5 were obtained.
... Vanadium metal has good performance under various temperatures, pressure and corrosive environmental conditions, which is one of the most sought rare metals, due to its special physical and chemical properties, such as corrosion resistance high hardness, high tensile strength, and high fatigue resistance [1][2][3]. At present, vanadium has been used in many fields such as aerospace [4], hydrogen storage devices [5][6][7], nuclear industry [8,9], superconducting alloys [10,11], etc. About 85% of vanadium is made into vanadium-containing iron-based alloys, VC, VN, etc. for use in the steel industry. ...
... Then Marden and Rich [23] produced metallic vanadium with a purity of 99% through a similar method firstly. V 2 O 5 , Ca and CaCl 2 were put in a closed container together, then heat to 900-950 • C for reaction and the reaction is as follow in Eq. (9). ...
Metallic vanadium has special properties and is widely used in metallurgy, aviation, nuclear industry and other fields. Vanadium is a rare and refractory metal, its high purity metal is difficult to prepare. The current mainstream extraction method is a combined process of aluminothermic reduction of vanadium oxide and electron beam smelting. Based on the properties of vanadium oxide and vanadium chloride, many researchers have also proposed various methods for the reduction of vanadium-containing precursors to prepare crude vanadium including metallothermic reduction, non-metallothermic reduction, electroreduction, and the refining of crude vanadium to produce high-purity vanadium such as molten salt electrorefining, iodide thermal decomposition, and solid state electrotransport. The review has carried out a more comprehensive summary of the various methods involved in the extraction of crude vanadium and the refining of crude vanadium, especially in recent years. The basic principles, development process, technical characteristics, and problems of corresponding methods are discussed in detail. It is expected to provide a reference for researchers and development of new extraction technologies of high-purity metallic vanadium.
... As is well known, the alloy composition and microstructure can be optimized by adding solute atoms in alloys design to improve their properties. In VeCreTi alloys, Cr is known to increase the high-temperature strength and Ti can enhance the ductility by absorbing interstitial impurities, such as C, O, and N [4]. However, excess Cr or Ti also has been reported to result in loss of ductility [4]. ...
... In VeCreTi alloys, Cr is known to increase the high-temperature strength and Ti can enhance the ductility by absorbing interstitial impurities, such as C, O, and N [4]. However, excess Cr or Ti also has been reported to result in loss of ductility [4]. Sakai et al. [5] reported that the ductileebrittle transition temperature of alloys increase with Cr concentrations, corresponding to the increase of flow stress due to solution-hardening by Cr. ...
In order to elucidate the role of solutes in vanadium alloys, the solute diffusions and the interactions of solute-solute and solute-vacancy have been investigated by means of the first-principles calculations. It is shown that large solutes are energetically favorable to combine with adjacent solute and to form the solute-solute nearest-neighbor pairs. In order to obtain qualitative understanding, the charge density differences for some special solute-solute binding are plotted to analyze the bonding interaction between solute atom and their nearest-neighbor host atoms. The solute-solute binding energies indicated that apart from the 3d transition metals, there is a clear trend of increasing binding energy with increasing solute volume. In particular, large solutes have larger binding energies for bonding with vacancies and can be considered as vacancy trappers in the crystal. As a comparison with solute diffusions, the self-diffusion coefficients of vanadium are determined. Activation energies for vacancy-mediated solute diffusion in vanadium are also determined. We conclude that the major contribution to the activation energy comes from the diffusion energy barrier, and both all follow the same trend. The noble metals have higher migration barriers than others, which means that these elements are inactive in the vanadium alloys. By contrast, the barriers for the large solutes almost cease to exist due to the formation of stable solute-centered divacancy.
... Therefore, vanadium alloys have been considered a promising candidate structural material for future fusion reactors. Among the typical vanadium-based alloys, V-4Cr-4Ti [4] alloy is noteworthy, where a Cr element is added to enhance high temperature strength and creep resistance, and a Ti element is added to improve the resistance to irradiation-induced void swelling [5]. However, the high sensitivity of interstitial impurities (such as oxygen, carbon, and nitrogen) limited the application and fabrication of vanadium alloys. ...
In the current work, an analysis of the effects of Y on the radiation hardening and microstructure response of a V-4Cr-4Ti alloy has been conducted after 30 keV D ion irradiation at room temperature using transmission electron microscopy (TEM) and nanoindentation. The results show that the formation of large Y2O3 and small Y2V2O7 nanoparticles was confirmed, indicating that the addition of Y reduces the amount of dissolved oxygen. The addition of Y has been shown to affect the radiation-induced dislocation loops, radiation hardening, and Ti-rich segregation of the V-4Cr-4Ti alloy. With the addition of Y, the mean size of the radiation-induced dislocation loop decreased, which may result from the strong sink strength of the nanoparticle/matrix interface, interactions between Y atoms and SIA clusters, and the strong binding energy of vacancy–oxygen pairs. Some particles with core–shell structures were observed after ion irradiation, where Ti-rich segregations at the nanoparticle/matrix interface were confirmed. These results indicate that Y might promote abnormal segregation. Possible causes for this include the lower interface energy at the particle/matrix interface and the interaction between oxygen and solute atoms.
... The spectrum of alloys is quite wide [96,97,329], for both fission and fusion [330] and especially considered for space nuclear applications [331], often combining elements, either in solid solution [96,97,329], or in layers [332,333]: molybdenum alloys [102,334,335] (U-Mo alloys are also considered as proliferation-resistant fuel [336].), niobium alloys [331,337], vanadium alloys [97,333,[338][339][340] and tungsten alloys [332,341,342], the two latter especially for fusion, and tungsten exclusively for it [329,342]. Tantalum alloys have been considered in the past in the US for space nuclear applications, but little information is available, except for a few conference abstracts that can be found online. ...
Nuclear energy is presently the single major low-carbon electricity source in Europe and is overall expected to maintain (perhaps eventually even increase) its current installed power from now to 2045. Long-term operation (LTO) is a reality in essentially all nuclear European countries, even when planning to phase out. New builds are planned. Moreover, several European countries, including non-nuclear or phasing out ones, have interests in next generation nuclear systems. In this framework, materials and material science play a crucial role towards safer, more efficient, more economical and overall more sustainable nuclear energy. This paper proposes a research agenda that combines modern digital technologies with materials science practices to pursue a change of paradigm that promotes innovation, equally serving the different nuclear energy interests and positions throughout Europe. This paper chooses to overview structural and fuel materials used in current generation reactors, as well as their wider spectrum for next generation reactors, summarising the relevant issues. Next, it describes the materials science approaches that are common to any nuclear materials (including classes that are not addressed here, such as concrete, polymers and functional materials), identifying for each of them a research agenda goal. It is concluded that among these goals are the development of structured materials qualification test-beds and materials acceleration platforms (MAPs) for materials that operate under harsh conditions. Another goal is the development of multi-parameter-based approaches for materials health monitoring based on different non-destructive examination and testing (NDE&T) techniques. Hybrid models that suitably combine physics-based and data-driven approaches for materials behaviour prediction can valuably support these developments, together with the creation and population of a centralised, “smart” database for nuclear materials.
... Vanadium-based alloys are considered candidates for the first-wall structure material of fusion reactors due to their low irradiationinduced activity, remarkable elevated temperature strength, and excellent resistance to radiation swelling [1,2]. In our previous work [7], we discovered that Ti-based precipitations are effective to strengthen Vanadium alloys. ...
In this work, the stacking fault energy (SFE), surface energy, and Rice ratio are systematically investigated to elucidate the effect of solute atoms (Cr, Ta, V, W, Y, and Zr) on the mechanical properties of TiC. As for the site preference of solute atoms, the calculation results show that all solutes prefer the Ti sites in the TiC. Considering the solute effect on the SFE, four slip systems, such as(001)<11¯0>,(110)<11¯0>,(111)<11¯0>, and(111) <112¯> are systematically investigated. For the undoped structure, we found that the (110)<11¯0> slip system is most likely to occur in the TiC due to its lower SFE, which is consistent with the experimental results. Furthermore, it is found that solute atoms Cr, V, W will decrease the SFE and improve the ductility of TiC, while doping Ta and Zr will increase SFE and reduce its ductility. Besides, it is worth noting that doping Y atoms not only decrease the stacking fault energy but also reduces the ductility of the material. Furthermore, results also show that all doped solutes are beneficial to the formation of partial dislocation in the TiC.
... Vanadium (V) alloys are considered as promising firstwall/blanket candidates for fusion reactors due to their lowinduced activation, high-temperature strength, excellent radiation tolerance, high thermal conductivity, compatibility with liquid lithium and so on [8][9][10][11][12][13] . In typical V-Cr-Ti alloys [ 14 , 15 ], as the alloying elements or low-concentration impurities, the substitutional solute atoms include the metal elements of Cr, Ti, Al, Fe, Mo, Nb, Cu, Ni, Ta and Zr and the nonmetallic elements of Si, S and P. The performance of V alloys could be enhanced by adding alloying elements, such as, the Cr and Ti additions can increase the hightemperature strength and radiation resistance and ductility, respectively. ...
The interactions of solute atoms with self-interstitial atoms (SIA)/clusters and their effect on stability of single SIAs and small SIA clusters in vanadium (V) were investigated using first-principles calculations. We calculated the binding energies of 16 substitutional solute elements (Cr, Ti, Al, Fe, Mo, Y, Nb, Cu, Mn, Ni, Ta, Zr, W, Si, S and P) with SIAs and found that the solute Cr/Fe/Mn/P/S prefers to form solute-V<111> mixed interstitial dumbbell. The interaction between solute Ti/Si/Y/Nb/Ta and <111> SIA dumbbell is attractive, while most of other solutes repel with the <111> SIA. The stable Cr-<111> mixed interstitials and the Ti-<111> SIA can attract five metal solutes (Ti, Y, Zr, Nb, and Ta), further stabilizing the SIA. The stability of small SIA clusters are determined and the parallel <111> SIA dumbbell clusters are the most stable configurations, followed by the parallel <110> SIA configurations. The small SIA clusters can attract three metallic solutes (Ti, Y and Zr) and two impurities (P and S), but will repel most solute atoms. The presence of solute atoms does not change the relative stability order among SIA clusters. The present calculations provide the basic data for understanding the influence solutes on the stability of SIAs/SIA clusters in V alloys under irradiation.
... Vanadium receives considerable attention due to its fast transcrystalline diffusion of hydrogen [23], favorable hydride phase stability, good thermal and mechanical properties, and corrosion resistance in environments such as PbLi [18]. In a nuclear environment, vanadium is preferred due to its lower induced radioactivity, which reduces high-level waste and decommissioning costs [24]. Therefore, for fusion applications, such as the tritium extraction system and MFPs, vanadium is the preferred material. ...
Dense vanadium-based membranes offer high permeability and perfect selectivity to hydrogen isotopes, maintain favorable neutronic properties, and are compatible with liquid metals such as PbLi. These properties make vanadium membranes a promising fusion fuel cycle technology for processes such as tritium extraction from PbLi and exhaust processing. Surface contamination has a deleterious effect on the gas-phase hydrogen permeation through vanadium, and the reported permeabilities range from 10⁻¹⁴ to 10⁻⁷ mol m⁻¹ s⁻¹ Pa-0.5. Thin dense films of palladium applied to clean vanadium surfaces enable a consistently high hydrogen permeability. In this study, uncoated vanadium resulted in deuterium permeabilities ranging from 2.8 × 10⁻¹¹ to 6.4 × 10⁻⁹ mol m⁻¹ s⁻¹ Pa-0.5 at 300 °C–700 °C, respectively. Post-test analysis revealed a VOx surface layer and VCx subsurface layer formed on the feed side, while the as-received surface oxide dissolved leaving a submonolayer oxide on the permeate surface. The Pd-coated V resulted in a maximum deuterium permeability of 2.1 × 10⁻⁷ mol m⁻¹ s⁻¹ Pa-0.5 at 375 °C upon activation of the Pd surface by oxidation and reduction. The deuterium permeation declined upon heating to 500 °C due to intermetallic diffusion between the Pd and V. The Mo2C-coated V resulted in deuterium permeabilities ranging from 2.7 × 10⁻¹⁰ to 1.8 × 10⁻⁹ at 500 °C–700 °C, respectively, and a post-test analysis found the carbon in the Mo2C layer had dissolved into the V near the interface.
... Thus, the requirements of this monumental engineering task demand great efforts in materials science research. Due to these stringent design requirements, only a handful of materials are considered as candidate structural materials for the first wall/blanket, chiefly among them vanadium (Muroga, 2012;Muroga et al., 2014;Muroga, 2017). Due to its superior nuclear performance and high-temperature mechanical properties, the V-4Cr-4Ti alloy family has been selected as the leading candidate for structural designs which employ liquid Li as a cooling agent (Le Flem, Gentzbittel, and Wident, 2013;Chen et al., 2011;Zinkle et al., 1998). ...
The emissivity is a thermophysical property that relates the amount of thermal radiation emitted by a material to that radiated by a blackbody. It is a crucial property for both industrial and scientific applications, since it determines the overall heat transfer in high-temperature or high-vacuum conditions. This thesis is divided into two main sections: on the one hand, the development and improvement of emissivity measurement methods at the University of the Basque Country (UPV/EHU) and, on the other, the application of these methods to the characterization of materials of industrial interest in the alternative energy sector (solar thermal energy and nuclear fusion power). Firstly, an in-depth review of the unique HAIRL emissometer has been carried out, including both instrumental and methodological improvements, as well as a renewed analysis of its sources of error. Secondly, three types of materials have been studied: multi-layer selective solar absorbers for parabolic trough power plants, non-selective black coatings for solar tower plants, and a family of vanadium alloys for future nuclear fusion reactors. The overall objective of this work is to improve the knowledge about the radiation heat transfer properties of key materials for alternative energy processes.
... En la siguiente figura se observa el comportamiento de un recubrimiento selectivo en función de la región espectral [1]. Otra de las aplicaciones para las cuales la radiación térmica es absolutamente crucial es la fusión nuclear (izquierda, [2]). En concreto, la primera pared del reactor (construida en aleaciones de V) debe ser capaz de disipar una gran cantidad de calor producido por el plasma, peró unicamente puede refrigerarse por radiación o Li fundido. ...
La radiación es uno de los tres modos de transferencia del calor, junto con la conducción y la convección. Se trata del proceso dominante a altas temperaturas, así como el único posible en condiciones de vacío. El estudio de la radiación térmica es fundamental para el desarrollo de numerosas aplicaciones energéticas a alta temperatura. El parámetro fundamental que lo caracteriza es la emisividad, que cuantifica cuánta radiación emite o absorbe un cuerpo comparado con la referencia teórica de un cuerpo negro. A lo largo de mi tesis doctoral, he analizado la respuesta emisiva de materiales diseñados para energías solar térmica y nuclear de fusión.
... Vanadium-based alloys are considered as promising first wall materials due to their low activation characteristics, remarkable high temperature strength and void swelling resistance in the fusion environment [1][2][3]. High number density of tiny precipitates dispersed in the matrix present the pinning effect on the dislocations glide and climb, and therefore improve the mechanical strength of V alloys. These nanoscale precipitates in V-4Cr-4Ti alloys are Ti-rich and most likely to be Ti-(O, N, C) with the NaCl structure [4]. ...
A correct description of the interaction between solutes and interface is the prerequisite for an understanding of the evolution and growth kinetics of precipitate. In this study, we use the first-principles calculations to characterize the solute (Cr, Ti, and Y) behaviors at the TiO-precipitate/V-matrix interface in Vanadium (V) alloys. After obtaining the equilibrium interface structure, the formation heat and segregation energies for solutes (Cr, Ti, and Y) are studied in detail to obtain the site preference and segregation behaviors for solutes in the interface region. We found that solute Ti prefer to the interface site and shows a segregation tendency at the interface. To the contrary, solute Cr and Y prefer to retain in the V matrix. However, from the standpoint of energy, the site preferences for all solutes are extremely weak when no vacancy is introduced to the interface. Considering the vacancy effect, the site preference of solute Y is changing and shows a segregation tendency at the interface. We also calculate the Griffith rupture work to uncover how the solutes influence the interfacial strength with or without the vacancy effect. Solute Cr is favorable for improving the interfacial rupture strength of the V alloys no matter where solute Cr is positioned. As for solute Ti and Y, when solutes are close to the interface, the interface-weakening effect is obvious. The improvement of solutes in interfacial strength is ascribed to the increased the hybridization behaviors between the solute atom and the Ti atoms in the precipitate.
... The ferrite-martensite 12% chromium steels EK-181(Fe-12Cr-2W-V-Ta), EP-823 (Fe-12Cr-W-V-Ni-Mo-Nb) and high-melting point low-activated vanadium alloy V-4Ti-4Cr are promising construction materials to be used in nuclear and fusion power reactors with various types of heat conductors (sodium, lead, lithium, lead-lithium) [1][2][3][4][5]. The mechanical and physical properties of these materials were studied in sufficient detail. ...
Static and fatigue strength at room temperature under conditions of repeated stretching of lowactivated ferrite-martensite 12% chromium steels EK-181 (Fe–12Cr–2W–V–Ta heat treatment + aging in lead at 600°C, 3000 h), EP-823 (Fe–12Cr–W–V–Ni–Mo–Nb, annealed condition), and alloy V–4Ti–4Cr (heat treatment + aging in lead at 600°C, 3000 h) were studied. It was shown that for materials there is straight-line dependence between the level of rupture resistance values and fatigue strength. The maximum fatigue limit of 600 MPa appears in steel EK-181 after a standard heat treatment and aging in lead at 600°C, 3000 h, and the minimal one of 300 MPa is observed in vanadium and V–4Ti–4Cr alloys. The fatigue failure mechanism is predominately of ductile character for all materials studied. The fatigue cracking originates near the surface and in some cases clustering of nonmetallic inclusions is the place of origin. The fatigue crack propagation is related to formation of a typical striation relief. Significant distinctions in the fracture surface relief of specimens after standard heat treatment and aging in liquid lead are not observed.
... E bind is the binding energies to form the corresponding VacO n complexes. Since the limited concentration for O, N, and C are lower than 1000 ppm in the typical V-4Cr-4Ti alloys, we assume that the initial concentration of O atom in the V alloys is 300 ppm which is the normal amounts of the O impurities in the high-purity V-4Cr-4Ti alloys [3]. Using Eq. (9), the predicted temperature dependence of the concentration of VacO n complexes are present in Fig. 3(c) it can be seen that all complexes concentrations increase significantly with temperature increases. ...
This study aims at characterizing the interstitial Oxygen (O) behaviors in the Vanadium (V) Alloy by means of the first-principles calculations. For this, the interations between vacancy (Vac) and O interstitil atom are studied in detail to obtain the binding energies and stable structures of the complexes. It can be seen that monovacancy binding with two O atoms occupied the opposing octahedral stie are particularly stable, and is liable to form VacO2 cluster in the V alloys. According to the mass action analysis, the predicted temperature dependence of the concentration for VacOn complexes are presented. Apart from monovacancy, we also consider the trapping behavior of vacancy cluster on the O atoms. The results also prove that one vacancy can trap two O atoms in the V alloys. Based the diffusion theory, we obtain the diffusion coefficients as a function of temperature with or without the vacancy effect in the V alloys. The predicted O diffusion coefficients in defect-free V alloys from our first-principles calculations are in excellent ageement with the experimental data, meanings that the vacancy-limited mechanism does not play the key role for O diffusion in V alloys. Regarding the interactions between vacancy, solutes and O atom, combining with the diffusion barriers of O affected by vacancy and solute, we infer the formation mechanism of the precipitates in the V alloys.
... Starting from 1970s Russia, Japan, and the USA actively develop low-activated radiation-resistant vanadium alloys for fusion power. By the 1990s the researches agreed that a first protective wall of a fusion reactor can be successfully made from V-4Ti-4Cr alloy [1][2][3][4][5][6]. ...
The development of low-swelling radiation-resistant alloys is vital for the creation of reliable fusion reactors. In this article, we revisit the long-standing problem of very low radiation swelling in V-Ti-Cr alloys by means of DFT calculations. In particular, we study double and triple interactions of point defects such as solutes, vacancies and self-interstitial atoms in bcc V. According to our results titanium atom and vacancy are strongly attracted and in addition to pairs form highly stable triple Ti-Vacancy-Ti complexes, which are absent in the case of chromium. By using an analytic model of void growth and using calculated binding energies of point defect complexes in bcc vanadium we obtain three orders of magnitude reduction of swelling rate due to the formation of Ti-Vacancy-Ti complexes, which allows to explain experimental observations. Finally, we explain the causes of the strong attraction between titanium and vacancy by means of geometry and electronic structure analysis.
... Vanadium alloys are considered as one of the most promising candidate materials for use in fusion reactors because of their low activation properties, as well as the high heat transfer capability, high-temperature operation, simple structure, high tritium breeding capability, and low tritium leakage. Major remaining issues of vanadium alloys are thermal and irradiation creep, transmutant helium effects on mechanical properties, and radiation effects on fracture properties (Muroga, 2012). ...
Developing new multifunctional materials in recent years for nuclear systems has become increasingly critical owing to the high demand on better shielding in extreme environments. The purpose of this research was to design, manufacture, and evaluate the feasibility of utilizing novel light weight close-cell composite metallic foam (CMF) and open-cell Al foam with fillers as radiation shields at nuclear facilities to attenuate the background of ionization radiations to a minimum level for creating a safer workplace, meeting regulatory requirements and maintaining high quality performance.
Steel-steel composite metal foams (S-S CMFs) and Aluminum-steel composite metal foams (Al-S CMFs) with various sphere sizes and matrix materials were manufactured and investigated for nuclear and radiation environments applications. 316L stainless steel, high- speed T15 steel and aluminum materials were used as the matrix material together with 2, 4 and 5.2 mm steel hollow spheres to manufacture various types of composite metal foams (CMFs). High-speed T15 steel is selected due to its high tungsten and vanadium concentration (both high-Z elements) to further improve the shielding efficiency of CMFs. This new type of S-S CMF is called High-Z steel-steel composite metal foam (HZ S-S CMF). Open-cell Al foams with fillers were obtained by infiltrating original empty pores with variety of hydrogen-rich compounds: petroleum wax, borated polyethylene, water, and borated water.
All the foams were investigated for their radiation shielding efficiency in terms of X-
ray, gamma ray and neutron. X-ray transmission measurements were carried out on a high- resolution microcomputed tomography (microCT) system. Gamma-emitting sources: 3.0mCi 60Co, 1.8mCi 137CS , 13.5mCi 124Am, and 5.0mCi 133Ba were used for gamma-ray attenuation analysis. The evaluations of neutron transmission measurements were conducted at the Neutron Powder Diffractometer beam facility at North Carolina State University. The experimental results were verified theoretically through XCOM and Monte Carlo Z-particle Transport Code (MCNP).
A mechanical investigation was performed by the means of quasi-static compressive testing. Thermal characterizations were carried out through effective thermal conductivity and thermal expansion analyses in terms of high temperature guarded-comparative- longitudinal heat flow technique and thermomechanical analyzer (TMA), respectively. The experimental results were compared with analytical results obtained from respectively Brailsford and Major’s model and modified Turner’s model for verification. Flame test was performed in accordance with United States Nuclear Regulatory Commission (USNRC) standard. CMF sample and a 304L stainless steel control sample were subjected to a fully engulfing fire with an average flame temperature of 800oC for a period of 30 minutes. Finite Element Analysis was conducted to secure the credibility of the experimental results.
This research indicates the potential of utilizing the lightweight close-cell CMFs and open-cell Al foam with fillers as shielding material replacing current heavy structures with additional advantage of high-energy absorption and excellent thermal characteristics.
... Low-activated vanadium alloys of V-Ti-Cr system (V-4Ti-4Cr is the best reference alloy) as the structural materials for cores of fusion and fission (fast) nuclear reactors are considered [1][2][3][4]. New compositions of vanadium low-activated alloys of V-Me(Cr, W)-Zr-С system demonstrate good prospects [5]. ...
Formation of regularities of the nanometric oxide precipitates and defect microstructure in vanadium-based low activation alloys V–Cr–Zr–(C,N,O) and V–Cr–W–Zr–(C,N,O) as a function of the regimes of their thermochemical treatment was investigated. Several methods of internal oxidation which provide formation of the nanosized ZrO2 particles of controllable dispersion, ensure the nanometric size of the heterophase structure to be maintained up to the temperatures as high as 1300–1400 °С, and allow the recrystallization temperature to be increased up to ≥1400 °С were proposed. Formation of such microstructure contributes to dispersion- and substructural hardening and results in more than twofold increase in the yield stress of these alloys both at room and elevated (800 °С) temperatures, compared to the conventional thermo-mechanical treatment.
... V-4Cr-4Ti alloy, which is a vanadium-based alloy containing Chromium (Cr) of 4 mass% and Titanium (Ti) of 4 mass%, is a candidate for blanket structural materials for fusion reactor systems because of its low-activation properties, good resistance to neutron radiation damage, and good high-temperature strength [1]. Nitrogen (N) and oxygen (O) have very high solubility limits of 1.7 mass% and 2.7 mass% at 1000 • C in vanadium, and are potent solid solution strengtheners [2,3]. ...
Reduction of interstitial impurities such as nitrogen (N) and oxygen (O) improves the mechanical properties of V-4Cr-4Ti alloys. Yttrium (Y) addition effectively reduces O content by Y2O3 slag-out on the melting ingot surface. The effects of Y addition on mechanical properties were investigated for V-4Cr-4Ti with N ranging from 0.009 to 0.29 mass% and O ranging from 0.009 to 0.36 mass%. The increase in yield stress (YS) and ultimate tensile stress (UTS) for vanadium alloys was saturated above 0.1 mass% in O content, because Ti precipitates increased with increasing O content regardless of Y addition. Y addition had little effect on YS and UTS of V-4Cr-4Ti alloys at room temperature (RT). However, Y addition improved impact properties of alloys highly doped with O. Y addition did not suppress hardening due to O doping but did increase deformation for crack initiation.
Among the wide range of natural uranium minerals, uranyl vanadates have attracted particular attention due to the economic viability of co-producing uranium and vanadium for industrial applications The typically remote locations of the ore deposits favour in situ analytical techniques offering rapid, minimally destructive characterisation wherever possible. This study reports on the use of luminescence, Raman and laser-induced breakdown spectroscopy to characterise the two most commonly found uranyl vanadates in nature, carnotite (K2(UO2)2V2O8·3H2O) and tyuyamunite (Ca(UO2)2V2O8·5–8H2O); the first attempt to use all three laser-based techniques in tandem for these phases. Significant differences in the luminescence emission signal intensity along with a noticeable shift in the resolved emission peak positions enable carnotite and tyuyamunite to be readily distinguished. Extraction of the band spacing from luminescence emission spectra provides confirmation of the position of the equivalent Raman uranyl symmetric stretch, ν1(UO2)²⁺, vibration for each mineral. Raman itself, was unable to differentiate carnotite and tyuyamunite owing to the weak signals obtained; however, the degree of splitting in the vanadate symmetric stretch, ν1(VO3)³⁻, feature and, to a lesser extent, the position of the ν1(UO2)²⁺ peak might be used to distinguish between tyuyamunite and metatyuyamunite, although further studies are required for confirmation. Compositional LIBS analysis was successful in identifying minor quartz, gypsum, Mg- and Fe-bearing inclusions, subsequently confirmed by optical and scanning electron microscopy. The findings indicate that multiple laser-based techniques offer the potential for real-time characterisation of uranyl vanadate phases where intrusive sampling, transportation and ‘off-site’ laboratory analysis is impracticable.
The vanadium alloys are attractive structural materials for applications in Generation IV reactors and fusion reactors. Surface protections of the alloy are required due to its high affinity to H, O etc. impurity atoms. The previous coating designs on the vanadium alloys have been focused on monolithic coatings and may have low coating-substrate bonding strengths. The TiAlN-based composite coatings which show remarkable surface protection effects, are yet to be applied to the vanadium alloy substrates. In the current work, a TiAl/TiAlN composite coating with intermediate TiAl layers was deposited on a V-4Cr-4Ti alloy substrate using the filtered-cathodic vacuum-arc-deposition (FCVAD) method. The coating with a total thickness of ∼18 μm was compact and free of any voids or inclusions. A high adherence strength of ∼82N was measured for the coating. The ultrafine equi-axed γ-TiAl grains in the TiAl layer, and the columnar grains in the TiAlN layer which are composed of a mixture of TiN/AlN/Ti3AlN phases were characterized in detail with the transmission electron microscopy (TEM). The surface hardness and surface roughness were significantly improved for the coating compared with the substrate. The electrochemical corrosion results also show that the composite coating offers improved corrosion resistance against aqueous corrosion at room temperature.
Corrosion and mass transfer phenomena that determine the choice of structural materials and operating parameters of the liquid metal coolant circuits are dealt in this chapter. The driving forces for the corrosion are first described. Since the phenomenon is unique for each coolant structural materials combination, the experimentally observed characteristics and the underlying mechanism in each case are presented: (1) oxygen-assisted general corrosion as well as carbon transport in sodium-steel systems, (2) corrosion of austenitic and ferritic steels and vanadium alloys in liquid lithium systems, (3) corrosion of steels in liquid lead and lead-bismuth eutectic alloy, and (4) corrosion of austenitic and reduced activity ferritic-martensitic steels in liquid lead-lithium eutectic alloy system. Methods adopted to mitigate corrosion in the lead and lead-bismuth eutectic systems, viz. control of dissolved oxygen in liquid metal, choice of composition of the steels, and microalloying of the surface with oxide formers such as aluminum are described. Since wetting of the base material by liquid metal is the prerequisite for corrosion and mass transfer to occur, observed characteristics of wetting by each liquid metal coolant are described.
The sputtering of vanadium particles at normal incidence wassimulated. The SRIM-code combined to a new ANGULAIR and SDTrimSPsimulation was employed to obtain the sputtering yields and the angular distribution of the atoms. The simulation was made for a large number of incident Kr+ions with 5 keV energy, letting the computer count the number of emitted particles in the solid angle. The angular distribution of differential sputtering yields of vanadium shows an over-cosine tendency.
Four V-Ta-4Cr-4Ti quaternary alloys containing different quantities of Ta were investigated to determine the effect of Ta content on the Charpy impact properties. Five button-shaped ingots of the V-4Cr-4Ti ternary alloy and V-xTa-4Cr-4Ti quaternary alloys (x = 3, 9, 15, and 22 wt.%) were fabricated on a laboratory scale by using non-consumable arc-melting in an argon atmosphere. Charpy impact tests were conducted at temperatures ranging from 77 K to 293 K using an instrumented impact tester. Both the upper shelf energy and the ductile-brittle transition temperature increased with increasing Ta content. The addition of 3 wt.% Ta resulted in solid solution strengthening without any degradation of the Charpy impact properties. Thus, the addition of 3 wt.% Ta (V-3Ta-4Cr-4Ti) is an appropriate amount to use in blanket structural materials for nuclear fusion reactors. The spectra of TEM-EDS for V-3Ta-4Cr-4Ti quaternary alloy indicate that there is no significant enrichment of Ta in the matrix as compared with that in the precipitate. However, thermal aging may result in the formation of the Laves phase, causing the degradation of Charpy impact properties. The characterization of precipitates, thermal aging, and creep tests of the V-3Ta-4Cr-4Ti quaternary alloy need to be investigated to determine the optimum Ta content.
We systemically investigated helium (He) interacting with 27 substitutional solute elements in vanadium by first-principles calculations. The interactions between early three 3d/4d/5d transition solutes and He are attractive while they are repulsive for other solute elements, He still favors tetrahedral sites. We predicted the influence of substitutional solutes on He effective diffusivity, Sc/Ti/Y/La can obviously reduce He diffusivity while other solutes show little effect. Moreover, we studied the synergistic interactions of the defect complex with He, solute and vacancy. The binding of group IIIB/IVB solution elements and vacancy reduce vacancy capturing for He, inhibiting He clustering in vacancies. The presence of He increases the stability of solute–vacancy clusters. The results can help to understand the effect of solute elements on He accumulation and embrittlement in vanadium alloys under irradiation.
Vanadium alloys are the promising first wall and blanket materials for fusion reactors. Large amounts of helium (He) and hydrogen (H) impurities are produced inside the materials along with irradiation defects under neutron irradiation, leading to bubble formation and microstructure changes, which will degrade the thermal and mechanical properties of vanadium alloys. The microstructure changes of materials are influenced by the interactions of point defects with solute atoms. Nowadays, first-principles calculations are intensively performed to elucidate these interactions, clustering, and dissolution, which can provide valuable information for the design of high-performance anti-irradiation materials. This paper reviews the recent findings of the interactions of point defects (vacancies, self-interstitial atoms) with substitutional solutes and interstitial solutes (C, O, N, H, and He) as well as their clusters in vanadium and its alloys from first-principles calculations.
First-principles calculations have been performed to investigate the interaction between solute impurity O and H/He/vacancy irradiation defects in Ti3AlC2. The formation energy and occupation of O atoms within different defects as well as the trapping progress of O/H clusters are discussed. It is found that O atom preferentially occupies the hexahedral interstitial site (Ihex-1) in bulk Ti3AlC2, whereas it prefers to occupy the neighbouring tetrahedral interstitial site (Itetr-2) within pre-exisiting Al monovacancy (VAl), Al divacancy (2VAl−Al) and the 2VAl−C divacancy composed of Al and C vacancies. The appearance of C vacancy could greatly reduce the oxygen formation energy and make O atom more inclined to occupy the center of C vacancy. Vacancy could capture more O atoms than H/He atoms, where the VAl and 2VAl−Al could hold up to fifteen and eighteen O atoms, respectively. Meanwhile, the O could also promote the formation of Al vacancy. On the other hand, O atoms tend to occupy the interstitial sites near Al atomic layer and have attraction to Al atoms, which is likely to enable O atoms to combine with Al atoms to form Al2O3 protective layer, thus effectively inhibiting further oxidation inside the Ti3AlC2. In addition, the H-O exhibits repulsion interaction, but strong attraction occurs in the He-O interaction. Therefore the O atom has an inhibitory effect on the formation of H cluster, while O could bind more He atoms to form a large number of He bubbles. Besides, the O impurity greatly reduces the trapping ability of vacancy to H atoms, and O and He have a synergistic interaction for inhibiting the aggregation of H clusters. The present results are expected to provide new insight into the behaviour of Ti3AlC2 under irradiation and oxidation conditions so that structural materials could be better designed.
This paper investigated the formation of bubbles and precipitates and the hardening effects in V–4Cr–4Ti alloy after irradiation of He and H ions and post-irradiation annealing. Microstructures observation shows that small bubbles formed in as-irradiated samples, and, during the post-irradiation annealing, lath-like precipitates formed and small bubbles grew upon. It was found that bubbles surrounding precipitates were significantly smaller than those away from, and tended to align along the precipitates. Composition analysis shows that the precipitates comprised TiO with FCC structure. The effects of bubbles and precipitates on hardening were evaluated by the nanoindentation test, as well as the dispersed barrier model that explicitly considered the small bubbles in as-irradiated sample, large bubbles and TiO precipitates in post-irradiated annealed samples. Compared with the as-irradiated samples, the higher hardness in post-irradiated annealed samples is attributed to the co-existence of bubbles and precipitates.
Vanadium (V) is sensitive to minute quantity of oxygen interstitials, which induce pronounced hardening and embrittlement. Here, we utilize oxygen to synthesize V solid solutions in order to reveal the mechanism of oxygen solutes induced hardening. With increasing of oxygen solute concentrations, the fracture modes of V samples transform from dimple, to a mixture of dimple and cleavage, and to a fully transgranular cleavage. High density of dislocations and dislocation debris are produced in strained samples. The mobility of screw dislocations is reduced and the dislocation cross-slip events are promoted by oxygen solutes. In addition to oxygen solution hardening, the generation of high density of oxygen-vacancy complexes plays a dominant role in the strengthening. High quantity of loop-shaped dislocation debris are direct evidence for the formation of oxygen-vacancy complexes. Profuse oxygen-vacancy complexes trap dislocations, promote cross-slips and assist dislocation storage, thus give rise to a superior combination of strengthening, strain hardening, and ductility in V with 1.0 at.% of oxygen. Once beyond a critical oxygen concentration (>1.6 at.%), V shows catastrophic brittle failure due to the exceptional high density of oxygen-vacancy complexes. These findings provide insight to design high performance refractory metals utilizing oxygen solutes.
The precipitates of second phases with coherent or incoherent interfaces can interact with dislocations to act as slip obstacles that can significantly improve the mechanics properties of alloys. In this work, the first-principles calculations based on density functional theory were performed to characterize the precipitate/Vanadium interface in the vanadium alloys and uncovering how structure of precipitations and interfacial defects (vacancy or solutes) distribution influence on the effectiveness of precipitation strengthening. Based the Baker-Nutting orientation relationship, the equilibrium stable precipitate/matrix interface structure are obtained. It is found that the so-called layer buckling exists in the interfacial region. Meanwhile, considering the equilibrium interface structure, the compressive strain is taken up by the Ti-based precipitation in the interface. For V-based precipitations, on the contrary, the interfacial strain is reversed that the tensile strain is taken up by the V-based precipitations. Under the uniaxial tensile loading, the stress-strain relations of heterostructures with different kinds of precipitates MX (M = Ti, V and X = C, N) are obtained to characterize the precipitation strengthening in the Vanadium alloys. The ideal tensile strength for heterostructure is sensitive to the size of precipitation. Furthermore, it is interesting to note that Ti-based precipitations not only improve the tensile strength, but also meliorate the ductility of the V alloys, which pointing to the inference that Ti-based precipitations are more effective to strengthen the V alloys than V-based precipitations. Our findings from this study can be implemented into providing the theoretical strategy for further design of the new style alloys.
The microstructure and the strength of high pressure torsion processed V–5Cr–5Ti alloy are investigated by X-ray line profile analysis and microhardness testing. High pressure torsion has been applied at 4 and 8 GPa with 0.25, 0.5, 1, 2, 4 and 8 rotations. The X-ray beam of the high angular resolution diffractometer, dedicated for line profile analysis, has a footprint of about 200 μm × 1.5 mm on the specimen. The diffraction patterns have been measured in the center, at half radius and close to the edge of the specimens. This technique has provided a large number of data of the microstructure and hardness as a function of strain up to about γ ≅ 300. The dislocation density determined by X-ray line broadening is correlated and discussed in terms of the strength of the alloy using the Taylor equation.
Vanadium alloys are attractive candidate structural materials in future fusion reactors. To evaluate the effects of irradiation on the deuterium retention behavior in vanadium alloys, samples made of V-5Cr-5Ti were irradiated by 5.5 MeV carbon ions. Doppler broadening measurements of the positron annihilation radiation tests were carried out to investigate the defect properties in the irradiated samples. Then the irradiated and virgin samples were implanted with deuterium in an ECR (electron-cyclotron resonance) linear plasma device followed by thermal desorption spectroscopy experiments. It was found that the irradiation process introduced large density of vacancy-type defects and subsequently increased the deuterium retention in the V-5Cr-5Ti, which remains as a serious concern for the vanadium alloys in the application as the structural materials of the fusion blanket.
The electrodeposition process of vanadium from LiCl–KCl base electrolytes was investigated by means of cyclic voltammetry, galvanostatic electrolyses and micro analytical analysis of the deposits. It is demonstrated that the valence state of the vanadium ions has a critical influence on the feasibility of performing a reproducible and stable coating process aiming to obtain compact vanadium films. When the electrolyte contained predominantly trivalent vanadium ions, the process was unstable and the deposit consisted of dendrites. In contrast, making use of a comproportionation reaction of metallic vanadium and VCl3 to divalent vanadium ions led to a stable deposition behaviour and allowed to obtain thick deposits with high current efficiencies. The disadvantageous behaviour of melts with mostly trivalent ions is explained by the fact that deposition is interfered by the reduction of trivalent to divalent ions under limiting current conditions.
Graphical Abstract
This study aims to characterize the interactions between substitutional solutes (3d, 4d and 5d transition metals) and interstitial impurities (C and O) in vanadium alloys, with or without the presence of an adjacent vacancy. For this purpose, the binding energies for solute-impurity and vacancy-impurity pairs, as well as solute-vacancy-impurity complexes are investigated by means of first-principles calculations, with or without the elastic correction. The vacancy-impurity binding energies suggest that it is energetically favorable to form stable 1nn vacancy-impurity pairs. For large-sized solutes, the solute-impurity interactions present strong repulsive interactions when a vacancy is absent, while showing strong attractive ones in the presence of a vacancy. Furthermore, a comprehensive study on the binding energy of defects revealed a positive correlation between the elastic correction energies and solute volumes, indicating that the elastic correction for the binding energies needs to be considered when a vacancy is absent in the vicinity of defects. Based on the binding preference, we can infer that a vacancy prefers to bond with large solutes adjacent to it and thus the resulting solute-vacancy pair can serve as a strong impurity trapper to form a defect complex, enhancing the nucleation and growth of precipitates in V alloys.
The processes of the accumulation and annealing of radiation-induced defects that occur under low-temperature (at 77 K) irradiation (with an energy E > 0.1 MeV) of V−4Ti−4Cr and V−10Ti−5Cr bcc alloys both nonmodified and modified with hydrogen isotopes in a concentration of 200 ppm, as well as the effect of these processes on the physicomechanical properties of these alloys, have been studied. It has been found that the saturation of these alloys with hydrogen leads to slight changes in their strength and ductility characteristics. The irradiation of the alloys at the temperature of 77 K results in a substantial increase in their yield stress and ultimate strength, as well as a decrease in their ductility. In the course of the postradiation annealing of the alloys at a temperature of 130 K, the stage related to the migration of interstitial atoms is observed. At temperatures of 290–320 K, the recovery stage occurs due to the formation of vacancy clusters. The stage that occurs at a temperature of 470 K can be attributed to the formation of impurity-vacancy clusters. Possible mechanisms of the radiation-induced strengthening of the alloys during irradiation and subsequent annealing have been discussed.
Vanadium alloys are advanced options for low activation structural materials. After more than two decades of research, V–4Cr–4Ti has been emerged as the leading candidate, and technological progress has been made in reducing the number of critical issues for application of vanadium alloys to fusion reactors. Notable progress has been made in fabricating alloy products and weld joints without degradation of properties. Various efforts are also being made to improve high temperature strength and creep-rupture resistance, low temperature ductility after irradiation, and corrosion resistance in blanket conditions.
Future research should focus on clarifying remaining uncertainty in the operating temperature window of V–4Cr–4Ti for application to near to middle term fusion blanket systems, and on further exploration of advanced materials for improved performance for longer-term fusion reactor systems.
Charpy impact tests of Japanese and US reference V-4Cr-4Ti alloys, NIFS-HEAT-1 (180 wppm oxygen) and US832665 (310 wppm Oxygen), were examined after gas-tungsten-arc (GTA) welding in a purified argon atmosphere. To investigate the effects of further reduction of oxygen level in the fusion zone, filler wires made of HP (high-purity V-4Cr-4Ti, 36 wppm oxygen) were used as well as those made of the reference alloys. Charpy impact property of NIFS-HEAT-1 in as-GTA-welded condition was superior to that of US832665. Use of the high-purity filler wires improved the impact property further. Good correlation was obtained between ductile-brittle-transition temperature (DBTT) and the oxygen level in the fusion zone. Since oxygen contamination from the atmosphere is avoidable by controlling its impurity level, oxygen reduction in weld materials, such as plate and wire, is crucial to obtain good weldability. Only contamination element detected in this study was hydrogen. Contamination occurred not only in fusion zone but also in base metal. Degassing of hydrogen after the welding may improve the impact property further.
Following the DEMO design analysis, two test blanket modules (TBM) were chosen in the RF for the development and testing in ITER: ceramic helium-cooled TBM and lithium self-cooled TBM. In the first one, lithium containing ceramics is used for tritium breeding, helium is used as a coolant and purge gas for tritium extraction, beryllium—as a multiplier. Ferritic steel is a structure material. In the second one lithium is used as tritium breeder and a coolant, and vanadium alloy of V–Cr–Ti system as a structure material.Conceptual designs of both TBMs and ancillary systems for their tests in ITER, strategy of tests, key R&D issues for both concepts are summarized. An international collaboration in R&D, development and testing of TBMs is of great importance due to shortage of testing space in ITER and due to high cost of the program.
Conceptual design studies for a tokamak-based demonstration fusion reactor have been carried out in Russia since 1991. The preferred concept was a steady-state operating tokamak with superconducting magnets, a single-null divertor configuration and a high contribution of bootstrap current into a plasma current drive. Two blanket concepts were analyzed: (1) a helium-cooled ceramic (Li4SiO4) design for tritium breeding, using ferritic steel as the structural material, and (2) a blanket using liquid lithium for the tritium breeding material and coolant and a vanadium–chromium–titanium alloy as the structural material. Conventional-type water/lithium-cooled divertor targets with a maximum heat load of ∼10MW/m2 were chosen. Blankets of both designs require beryllium as a neutron multiplier and have to be replaced after the integral fusion neutron load on the first wall reaches 10MWa/m2. The results of the analyses show the necessity of additional studies prior to choosing the most promising blanket concept for further development. Aspects of radioactive waste management and scarce material recycling were also considered.
Pressurized thermal creep tubes of V–4Cr–4Ti have been examined following testing in the range 650–800°C for tests lasting ∼104h to provide comparison with tests on similar tubes following irradiation. It is found in all cases that creep results from dislocation motion. But the mechanism changes with increasing temperature and lower stress from one controlled by the climb and interaction of individual dislocations, to one controlled by sub-grain boundary structure that is created by relaxation of the interacting dislocations to lower energy planar arrays. This change in mechanism corresponds to a change from power law creep to Newtonian creep such that the stress exponent drops from ∼4 to ∼1. Although it is possible to explain the Newtonian response as Nabarro–Herring or Coble creep, it appears more likely that behavior is due to Harper–Dorn creep, in which case the change in response occurs at the Peierls stress.
The paper presents irradiation creep data for V–4Cr–4Ti irradiated to 3.7dpa at 425 and 600°C in the HFIR-17J experiment. Creep deformation was characterized by measuring diametral changes of pressurized creep tubes before and after irradiation. It was found that the creep strain rate of the US Heat 832665 of V–4Cr–4Ti exhibited a linear relationship with stress up to ∼180MPa at 425°C with a creep coefficient of 2.50×10−6MPa−1dpa−1. A linear relationship between creep rate and applied stress was observed below ∼110MPa at 600°C with a creep coefficient of 5.41×10−6MPa−1dpa−1; non-linear creep behavior was observed above ∼110MPa, and it may not be fully accounted by invoking thermal creep. The bilinear creep behavior observed in the same alloy irradiated in BR-10 was not observed in this study.
The effect of V–(0–70)Ti–(0–30)Cr (at.%) compositions on their compatibility with nitrogen-containing lithium (0.0015–0.67 at.% N) at 7000°C under steady-state test conditions and long-term contact with lithium (up to 2000 h) has been studied. The conditions for formation and stable coexistence of nitride layers on the surface of various compositions under variable nitrogen concentration in lithium have been defined. The V–(8–10)Ti–(4–5)Cr compositions showed the best characteristics from the standpoint of corrosion resistance, nitride layer stability under conditions of variable nitrogen concentration in lithium, and the possibility of ‘in situ’ protective nitride layer formation.
Tungsten (W) coating on fusion candidate V-4Cr-4Ti (NIFS-HEAT-2) substrate was demonstrated with plasma spray process for the purpose of applying to protection of the plasma facing surface of a fusion blanket. Increase in plasma input power and temperature of the substrate was effective to reduce porosity of the coating, but resulted in hardening of the substrate and degradation of impact property at 77 K. The hardening seemed to be due to contamination with gaseous impurities and deformation by thermal stress during the coating process. Since all the samples showed good ductility at room temperature, further heating seems to be acceptable for the vanadium substrate. The fracture stress of the W coating was estimated from bending tests as at least 313 MPa, which well exceeds the design stress for the vanadium structure in fusion blanket.
Because of the high solubility and mobility of oxygen in vanadium, composition control during the fabrication of thin (0.25 mm) wall tubing from vanadium alloys by cold drawing and annealing, presents a technological challenge. During intermediate annealing at 1000 °C in the 10-4 Torr vacuum regime, oxygen penetration into the tube wall is controlled by the development of a semi-protective surface oxide (linear-parabolic oxidation conditions); oxygen-hardened surface layers lead to a high incidence of surface cracking during the final stages of cold drawing. In the 10-5 Torr regime, under linear kinetic oxidation conditions, rapid oxygen penetration results in unacceptably high levels of oxygen pick-up (˜1500 wppm). In the 10-7 Torr vacuum regime, molecular impingement rates are reduced to the point where overall oxygen pick-up is reduced to
A mono-metallic V–4Cr–4Ti thermal convection loop was operated in vacuum (∼10−5Pa) at a maximum Li temperature of 700°C for 2355h and Li flow rate of 2–3cm/s. Two-layer, physical vapor deposited Y2O3–vanadium, electrically insulating coatings on V–4Cr–4Ti substrates as well as tensile and sheet specimens were located in the flow path in the hot and cold legs. After exposure, specimens at the top of the hot leg showed a maximum mass loss equivalent to ∼1.3μm of metal loss. Elsewhere, small mass gains were observed on the majority of specimens resulting in an increase in hardness and room temperature yield stress and a decrease in ductility consistent with the observed uptake of N and C from the Li. Specimens that lost mass showed a decrease in yield stress and hardness. Profilometry showed no significant thickness loss from the coatings.
Fusion reactors have been proposed with a vanadium alloy as the structural/containment material. However, vanadium has a significant affinity for interstitial contamination that could deleteriously affect its mechanical properties. The effects of oxygen pick-up in air and low pressure oxygen environments were investigated at 400–500°C for two VCrTi alloys. As expected the studies showed that the room temperature tensile ductility is reduced by exposure to air or low pressure oxygen environments. However, the magnitude depends upon processing history and subsequent heat treatment. Possible embrittling mechanisms such as grain boundary weakening or weakening of near-boundary regions are discussed.
An assessment of the issues on using flibe for fusion applications has been made. It is concluded that sufficient tritium breeding can be achieved for a flibe blanket, especially if a few cm of Be is include in the blanket design. A key issue is the control of the transmutation products such as TF and F{sub 2}. A REDOX (Reducing-Oxidation) reaction has to be demonstrated which is compatible to the blanket design. Also, MHD may have strong impact on heat transfer if the flow is perpendicular to the magnetic field. The issues associated with the REDOX reaction and the MHD issues have to be resolved by both experimental program and numerical solutions.
The mechanical properties of V–Cr–Ti type alloys depended on heat treatment conditions and Cr concentrations. In this paper the correlation between mechanical properties and heat treatment conditions as a function of Cr concentrations was explored using mini-size Charpy impact tests and microstructure observations. Vanadium alloys evaluated were V–xCr–4Ti (x=4, 7, 10, 12, 15, 20). Microstructure observations using transmission electron microscope (TEM) were performed to characterize the precipitates. The typical precipitates in the V–xCr–4Ti alloys were identified as Ti(C,O,N) and TiO2. Ti(C,O,N) was observed in specimens annealed between 900 and 1000 °C and TiO2 was between 1100 and 1200 °C. The DBTT of V–(4,7)Cr–4Ti alloys was around −190 °C, while DBTT for the other alloys were above −30 °C. It is proposed that formation of precipitates larger than 400 nm in diameter found in the alloys containing more than 10% of Cr as well as solution hardening of Cr are affected to the increase of the DBTT of the alloys.
The deuterium retention properties of vanadium alloy, V–4Cr–4Ti, were investigated by thermal desorption spectroscopy after deuterium ion irradiation. The deuterium ion irradiation was carried out at 380, 573 and 773 K with ion energy of 1.7 keV. Deuterium retained in the sample was desorbed in the forms of D2, HD, HDO and D2O. The amount of retained deuterium at 380 K increased with the ion fluence, and did not saturate to a fluence up to 1 × 1019 D/cm2. In addition, more than 100% of implanted D was retained at high deuterium fluence. The retained amount at 380 K was one and two orders of magnitude larger than graphite and tungsten, respectively. For the irradiation at 773 K, the amount of retained deuterium decreased with increase in ion fluence in the high fluence region, and the retained amount was almost the same as that for graphite or tungsten.
The corrosion of V/20wt%Nb/10wt%Mo, V/20wt%Nb/5wt%Cr, and V/20wt%Ti alloys exposed to high-purity liquid sodium has been examined. The oxygen absorption, the dissolution of solid metals into the sodium, and the deterioration of mechanical properties of the alloys were suppressed by decreasing the oxygen concentration in the sodium. The changes of tensile properties of the V/20wt%Nb/10wt%Mo and V/20wt%Nb/5wr%Cr alloys exposed to the sodium are interpreted in terms of oxygen in solid solution, and those of the V/20wt%Ti alloy in terms of internal oxidation of titanium in the alloy. The 700°C tensile elongation values of the V/20wt%Nb/10wt%Mo and V/20wt%Nb/5wt%Cr alloys were more than 15 and 20 percent, respectively. The room-temperature ductility of the three alloys was substantially reduced.
Specimens of V–4Cr–4Ti alloy were heated at 1273 K in vacuum, and the influence of this heat treatment on H2 absorption was examined at temperatures from 523 to 1023 K under the presence of water vapor of 10−5 Pa. The rate of H2 absorption was significantly reduced by the heat treatment in the temperature range examined. Such reduction in the absorption rate was ascribed to the surface segregation of Ti and increase in surface oxygen coverage caused by preferential oxidation of segregating Ti by water vapor. Comparison with data reported by other researchers [J. Nucl. Mater. 233–237 (1996) 376; Fusion Technol. 34 (1998) 868; J. Nucl. Mater. 233–237 (1996) 510] indicated the strong barrier effect of Ti oxide against hydrogen ingress.
Exposure of V–Cr–Ti alloys to low oxygen partial pressures at high temperature results in oxygen absorption and internal oxidation. Characterization of a V–4Cr–4Ti alloy after oxidation at 500°C revealed a microstructure with ultrafine oxide precipitates in the matrix and along grain boundaries. Heat treatment at 950°C following oxidation resulted in large TiOx precipitates in the matrix and grain boundaries. Tensile ductility was reduced by exposure to low-pressure oxygen under the temperature and pressure conditions. However, heat treatment at 950°C following oxidation was generally effective in recovering ductility irrespective of initial annealing treatment or grain size. Without increases in oxygen, >500 wppm hydrogen was required to cause significant decreases in tensile elongation. When oxygen was added either during or prior to hydrogen exposure, significant embrittlement occurred with 100 wppm hydrogen. Because of this synergism with hydrogen, oxygen pick-up remains a major concern for V–Cr–Ti alloys in fusion reactor applications.
Pressurized thermal creep tubes of highly purified V–4Cr–4Ti, the NIFS-Heat-2 alloy have been examined following testing in the range 700–850 °C. It was found that the creep stress exponent of the NIFS-Heat-2 alloy is about five and that the characteristic creep mechanism was the dislocation creep usually observed in pure metals. The apparent activation energy of creep deformation is about 210 kJ/mol in the temperature range 700–850 °C. Creep deformation was considered to be controlled by climb-controlled dislocation-glide at 850 °C, where sub-grain boundary structure predominates and consists of dislocation dipole structures and pile-ups of dislocations.