Search

V-Cr-Ti alloys with reduced Ti contents are expected to be good candidates as recyclable structural materials for fusion reactors. Herein, He irradiation was applied to V-4Cr-xTi (x = 0 to 4) alloys to investigate the additional effect of Ti and gas interstitial impurities on their microstructural evolution and irradiation hardening. Following He irradiation of the specimens at 700 °C with a maximum damage depth of 0.5 dpa, transmission electron microscopy and nanoindentation hardness tests were conducted. The microstructural evolution of the titanium-oxycarbonitride (Ti(CON)) precipitates formed during He ion irradiation was observed. The behavior of irradiation hardening is explained by the formation process of Ti(CON) precipitates and the magnitude of irradiation hardening depends on the amount of gas interstitial impurities that contribute to precipitate formation.

Vanadium alloys are an attractive candidate material for advanced fusion reactors’ structural components. Some leading vanadium alloys, such as V−(4−5)Cr−(4−5)Ti alloy, exhibit several important advantages, including excellent strength at elevated temperatures, high resistance to neutron irradiation damage, inherently low long-term activation, as well as good fabricability and weldability. However, the corrosion and embrittlement via oxygen pickup during the high-temperature oxidation process of vanadium alloys remains a key issue, restricting their operation conditions and long service life. In a high-pressure oxygen environment, the main oxidation product V2O5, with a low melting point of ～680 ℃, is formed on the vanadium alloy surface, which cannot offer reliable protection to mitigate further oxidation over 650 ℃. However, despite being exposed to a very low-pressure oxygen environment, it is still unlikely for vanadium alloys to form an effective oxidation film to retard the oxygen absorption at temperatures over 450 ℃, mainly due to the high solubility of oxygen in vanadium. When the oxygen concentration reaches 0.2% in the matrix of V−4Cr−4Ti alloy, it can cause severe oxygen embrittlement, possibly due to oxygen accumulation and formation of fine oxidation precipitates at the grain boundaries and the adjacent matrix. Therefore, it is significantly important to enhance the high-temperature oxidation-resistant performance of the vanadium alloy to broaden the operation conditions. In this work, this research progress on the high-temperature oxidation resistance of vanadium alloys is systematically reviewed. In summary, three main methods for enhancing the oxidation–corrosion resistance of vanadium alloys at elevated temperatures are elaborated, i.e., oxidation-resistant element addition, diffusion coating, and overlay coating. Additionally, the characteristics and existing problems of these methods and the responding examples are also analyzed and discussed in detail. In the first two methods, it is impossible to completely isolate the alloy substrate from the service environment; thus, the typical oxidation product V2O5 is easily formed in the high-pressure oxygen environment, leading to severe oxidation corrosion and embrittlement, especially at elevated temperatures. Expectedly, the dense overlay coating presents a greater potential application mainly because of the thorough protection from the service environment. Finally, the development trend in the modification and technical requirements of the advanced overlay coatings on high-performance oxidation resistance are prospected in this paper as per the practical application demands for vanadium alloys, aiming to provide a beneficial reference for further research.

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. [ABSTRACT FROM AUTHOR]

V-based alloy membranes have huge potential for hydrogen separation and purification due to their higher hydrogen permeability, but suffer from hydrogen embrittlement. To address this matter, Pd and Fe have been chosen as alloying additions in V-based membranes to inhibit hydrogen embrittlement and maintain high hydrogen permeation performance based on the lattice-matching strategy. The phase structure, composition, microstructure and hydrogen permeation properties of V-Pd-Fe ternary alloy are investigated systematically, and the V-Fe and V-Pd binary alloys are presented for comparison. The results show that the prepared V-Pd-Fe ternary can match almost perfectly with pure V in lattice constants. Attributing to the lattice-matching strategy, the V-Pd-Fe ternary alloy exhibits more excellent hydrogen permeable properties than that of V-Fe and V-Pd binary alloys. The reduction of hydrogen embrittlement is ascribed to the lower hydrogen solubility that induced by the lattice distortion and electronegativity of Fe and Pd. Meanwhile, the manipulated lattice constant that parallel to pure V maintains high hydrogen permeation performance. Additionally, the designed ternary V 87.8 Pd 8 Fe 4.2 alloy exhibits exceptional long-term hydrogen permeation stability at 573 and 623 K, and hydrogen flux gradually decreases at 673 and 723 K is observed due to the interdiffusion between Pd film and V 87.8 Pd 8 Fe 4.2 substrate and defects in the Pd coating. This study offers new scenario and method for optimized hydrogen permeable properties and restrains hydrogen embrittlement in the alloy membrane. • Constructing the lattice structure of V-Pd-Fe alloy by modulating Pd and Fe composition is almost identical to that of pure V. • The ternary V-Pd-Fe alloy exhibits excellent hydrogen permeability proprieties compared to alloys of binary V-Pd and V-Fe. • The mechanism for H solubility reduction and long-term H permeation deterioration in V-Pd-Fe alloys are studied. [ABSTRACT FROM AUTHOR]

The irradiation behavior of vanadium alloys was examined to evaluate the effect of temperature history during reactor operation. Irradiation experiments were performed in a fast reactor, Joyo, with an irradiation-temperature-controlled irradiation rig, MARICO-II. In-core irradiation rigs that control the irradiation temperature and are not influenced by reactor power were used. Irradiation conditions with or without the MARICO-II rig were performed. Significant differences in irradiation hardening and swelling behavior in undersized vanadium binary alloys were visible between temperature-controlled and conventional rigs. The temperature history affects the microstructural evolution, especially the nucleation and growth of voids during startup and shutdown procedures. No difference was observed in the vanadium alloys with oversized solute atom because the vacancy migration was suppressed by tight binding between oversized solute atoms and vacancies during irradiation. V–4Cr–4Ti–(Si,Al,Y) alloys showed significant irradiation hardening at high-temperature irradiation even though the contribution of dislocation and precipitates for irradiation hardening was insufficient to satisfy the measured irradiation hardening. The significant irradiation hardening in the V–4Cr–4Ti–(Si,Al,Y) alloys may be caused by transmission-electron-microscope-invisible clusters including minor elements (Si, Al, Y) that were formed at a high temperature during irradiation.

Intermetallic compounds often have a low radiation tolerance due to the low recombination rate of radiation-induced defects. In the present work, we designed a novel multicomponent vanadium-based alloy (MVA), V34Ti25Cr10Ni30Pd1, containing a micron-scale TiNi matrix phase with VCr nanoprecipitates (NPs) and a micron-scale VCr matrix phase with TiNi NPs. The MVA was irradiated with 6 MeV Ti3+ ions with a radiation dose of 5×1015ions/cm2 at room temperature. Results indicated that the micron-scale intermetallic TiNi matrix along with VCr NPs inside both became amorphous, while the micron-scale VCr matrix including numerous intermetallic TiNi NPs both exhibited a high structural stability after ion irradiation. The intermetallic TiNi matrix becomes amorphous due to the accumulation of radiation-induced defects, and the intermetallic TiNi NPs in VCr matrix have a high stability of crystallographic structure due to high-density interfaces between NPs and matrix. These results indicate that the phase stability of TiNi intermetallic compound is increased after nanocrystallization. Besides, the discrepancy of two matrix phases including precipitates after ion irradiation, and the underlying mechanisms are discussed in detail in this work. This work gives a guidance for designing new vanadium alloys and intermetallic compounds with enhanced structural stability under ion irradiation.

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.

Copyright of Journal of Central South University is the property of Springer Nature and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Activation analysis of impurities was performed on a test specimen of the reference low-activation vanadium alloy NIFS-HEAT-2 (V-4Cr-4Ti) which has been irradiated in Japan Materials Testing Reactor (JMTR) ~16 years before. Since the radioactive nuclides from V, Cr and Ti are produced with short half-lives, intense peaks of 60Co (T1/2 = 5.27 years) were dominant in the measured gamma-ray energy spectra and peaks of 94Nb (2.03 × 104 years) were also observed. The concentrations were evaluated to be 0.41 wppm and 0.71 wppm respectively from comparison with results of radioactivity calculation using the FISPACT-2005 code. In the analysis, detection of 108mAg (418 years), which would have a considerable influence on the material recycling after the usage in a fusion reactor, was intense motivation. Although the peaks of 108mAg has not been detected, the concentration of Ag was estimated to be

An on-line instrumented material irradiation rig (MARICO-II) was used for investigating temperature history effect for void swelling behavior in V-Fe binary alloys irradiated at a fast rector, Joyo. Neutron irradiation were performed at 477C with 7.5 dpa in the temperature control capsule (TC capsule) and at 482C maximum temperature with 20.4 dpa in the conventional capsule with monitoring temperature. In the TC capsule, voids was formed and distributed homogeneously, on the other hand a lot of tiny void formation or bimodal void distribution were observed in the conventional capsule, The void swelling rate of V-Fe alloys irradiated in the TC capsule was larger than that in the conventional capsule and it is caused by the imbalance of point defect flux to sink due to void nucleation at the low temperature transient regime in start-up and shutdown period of each cycle in neutron irradiation, that is temperature history effect for swelling behavior of V-Fe binary alloys during neutron irradiation.

• Ne ion irradiation produced significant hardening than Fe ion irradiation in V alloy. • The significant hardening produced by Ne ion irradiation contributed to the formation of more dislocation loops rather than bubbles with limited sizes. • The presence of inert gas atoms had a thermal stabilization effect on irradiation hardening with a required critical atom concentration. Vanadium (V) alloys are attractive candidate materials for tritium breeding modules (TBM) of blankets in fusion reactors. However, irradiation-induced hardening and embrittlement give rise to significant challenges for the application of V alloys in advanced nuclear reactors, particularly in the presence of inert gas atoms. In this study, V-5Cr-5Ti alloys were irradiated by 122 MeV Ne ions (at doses of 0.1 and 0.3 dpa) and 352 MeV Fe ions (at doses of 0.06, 0.20 and 0.35 dpa) at a temperature around 100 °C, and subsequently annealed at the temperature ranging from 212 to 500 °C. During the annealing process, the hardening produced by Fe ion irradiation started to recover at a temperature higher than 212 °C and was eliminated at a temperature higher than 375 °C. For the sample with Ne ions irradiation, the irradiation hardening of the sample with 0.3 dpa can be sustained at a temperature up to 450 oC, but an evident reduction of hardening in the sample with 0.1 dpa was observed after annealing at 212 oC. This suggested that the existence of inert gas atoms had a thermal stabilization effect on irradiation hardening, but it required a critical atom concentration. The different annealing behaviors of irradiation hardening at different doses under Ne ions irradiation were considered to be from the coupling effect of bubble growth and the interaction between interstitial clusters and Ne atoms. [ABSTRACT FROM AUTHOR]

V–4Cr–xTi (x = 0 to 4) alloys were used to investigate the additional effect of Cr, Ti and interstitial impurities on the microstructural evolution in He-irradiated V–Cr–Ti alloys to minimize radioactivity after fusion neutron irradiation. Transmission electron microscopy and atom probe tomography were carried out to the He-irradiated specimens at 500 °C with 0.5 dpa at peak damage. A flash electro-polishing method for the FIB-extracted specimen was established for the ion-irradiated vanadium alloys. The microstructural evolution of the irradiation-induced titanium-oxycarbonitride, Ti(CON) precipitates was observed and was influenced by the effect of Ti addition on the Ti(CON) precipitation. Apparent Ti(CON) precipitates formed in V-4Cr-xTi with 2% addition of Ti. In the V-4Cr-1Ti alloy, a high density Ti enriched cluster was formed. The origin of the irradiation hardening increase resulted from the size distribution of Ti(CON) precipitation from the dispersed barrier-hardening theory.

Microstructure and irradiation hardening of V-4Cr-4Ti alloys with different yttrium (Y) contents were studied by self-ion irradiation at 550 °C via TEM and nano-indentation test technology. The peak damage of the V-4Cr-4Ti-xY alloy (x = 0, 0.1, 0.5, and 1, wt.%) irradiated by 2.5 MeV self-ion (V2+) is 8 dpa. Dense dislocation loops were observed in all vanadium alloy samples after irradiation. With the increase of Y content, both average size and number density of dislocation loops using g = near the pole [001] decreased, while the irradiation hardness increment first decreased and then increased. In order to better reduce the irradiation hardening, it is considered that the addition of 0.1 wt.% Y in V-4Cr-4Ti alloy is reasonable.

TEM observation and tensile test were examined for vanadium alloys irradiated in a temperature control rig in JMTR at 290°C with damage level ranged from 0.003 to 0.06dpa. With the increase of the neutron dose, irradiation hardening could be observed in all the vanadium alloys except for the V–5Nb alloy. In the case of pure vanadium, the relationship between irradiation hardening and neutron dose was described as Δσ ∝ (ϕt)0.35-0.53. For V–5Cr alloy and V–4Cr–4Ti–0.1Si alloy, the dose dependence on irradiation hardening increase was shown as Δσ ∝ (ϕt)0.8 and Δσ ∝ (ϕt)0.8-1.0, respectively. From the TEM observation, the hardening source of radiation-induced defects was mainly determined to be dislocation loops for pure vanadium, loops with voids for V–5Cr and, loops and {100} precipitates for V–4Cr–4Ti–0.1Si and V–3Fe–4Ti–0.1Si alloys. From the strain rate dependence of 8% stress for V–4Cr–4Ti–0.1Si alloys tested at RT, the strain rate sensitivity, m=1/σ*(dσ/dln(dε/dt)) shows positive. Therefore, the dynamic interaction between interstitial impurities and dislocation is not strong in V–4Cr–4Ti alloys in the temperature range from RT to 290°C. A discrepancy of deformation mode of irradiated V–4Cr–4Ti–0.1Si alloys with 0.068dpa could be seen when the charpy impact test indicated the brittle behavior and the tensile test indicated the ductile behavior at room temperature. It can be explained by the difference of strain rate for the value of yield stress between tensile test and charpy test and the critical fracture stress.

A set of V–(4–8)Cr–(0–4)Ti alloys was fabricated to survey an optimum composition to reduce the radioactivity of V–Cr–Ti alloys. These alloys were subjected to nano-indenter tests before and after 2-MeV He-ion irradiation at 500 °C and 700 °C with 0.5 dpa at peak damage to investigate the effect of Cr and Ti addition and gas impurities for irradiation hardening behavior in V–Cr–Ti alloys. Cr and Ti addition to V–Cr–Ti alloys for solid–solution hardening remains small in the unirradiated V–(4–8)Cr–(0–4)Ti alloys. Irradiation hardening occurred for all V–Cr–Ti alloys. The V–4Cr–1Ti alloy shows the highest irradiation hardening among all V–Cr–Ti alloys and the gas impurity was enhanced to increase the irradiation hardening. These results may arise from the formation of Ti(CON) precipitate that was produced by He-ion irradiation. Irradiation hardening of V–Cr–1Ti did not depend significantly on Cr addition. Consequently, for irradiation hardening and void-swelling suppression, the optimum composition of V–Cr–Ti alloys for structural materials of fusion reactor engineering is proposed to be a highly purified V–(6–8)Cr–2Ti alloy.

Abstracts First-principles calculations are performed to derive promising alloying elements which inhibit inter-diffusion between vanadium and palladium-coating layers in vanadium-based hydrogen separation membranes. Paying attention to the inhibition of inter-diffusion by the grain boundary segregation of impurities, several physical features of various solute elements, grain boundary segregation, grain boundary embrittlement and solute–solute binding tendencies, are evaluated to construct a large database. Several alloying elements and their combinations similar to or even better than a previously reported yttrium are suggested for future experimental works by comprehensively analyzing these physical features. [ABSTRACT FROM AUTHOR]

The hydrogen solubility in a FLiNaK molten salt mixed with Ti powder was investigated. A hydrogen-soluble metal powder mixed with a molten salt can increase the effective hydrogen solubility of the molten salt, which is currently a major disadvantage of molten salts. A significant increase in hydrogen solubility was observed, even with a mass fraction of Ti powder of only 0.1 wt%. The increase of hydrogen solubility was so large that a vanadium alloy would be compatible with the Ti powder/molten salt mixture, unlike typical molten salts that result in an unacceptably large tritium inventory in the vanadium alloy. In addition, contamination of the Ti powder by oxidation suppressed the hydrogen uptake and release capability. Controlling the purity of the molten salt was found to be one of the key issues for the metal powder mixture concept. [ABSTRACT FROM AUTHOR]

Highlights: [•] Addition of 10mol% Cr to V increases the β-hydride T C to >200°C. [•] Addition of 10mol% Ni to V increases the β-hydride T C to >400°C. [•] Addition of 10mol% Al to V decreases the β-hydride T C to <30°C. [•] V90Al10 membrane can be cycled to <30°C under H2 without β-hydride formation. [ABSTRACT FROM AUTHOR]

Abstract: We investigate the metal–hydrogen interaction in vanadium alloyed with X=Fe, Ni, Al, Ti, and Nb using first-principles calculations. The site energy of interstitial hydrogen depends on both the volume and the presence of a nearby alloying element; the two effects are analyzed separately. The effective volume of the alloying element follows the sequence Nb>Ti>Al>V>Ni>Fe, and a larger volume lowers the site energy on average. Locally, Al appears to be the strongest hydrogen repellent, and Ti exhibits the strongest affinity for hydrogen. The original tetrahedral coordination by the metal atoms is also affected and the effect is the most pronounced in Ni, where the hydrogen atom in the octahedral interstitial site is strongly stabilized. A 16-atom body centered cubic special quasi-random structure is utilized to obtain the statistical distribution of site energies in V12X4. The obtained site energies at 673K have the following sequence: Ti

Abstract: W coating by vacuum plasma spray process and Cr coating by chromizing process were performed on fusion low activation materials, JLF-1 ferritic steel and NIFS-HEAT-2 vanadium alloy. The present study discusses feasibility of the coatings as anti-corrosion coating against fluoridation in Flibe for fusion low activation materials. Coatings were characterized by microstructural analysis and examination on chemical stability by corrosion tests. The corrosion tests were conducted with H2O–47% HF solution at RT and He–1% HF–0.06 H2O gas mixture at 823K to simulate fluoridation and oxidation in Flibe. The coatings presented suppression of fluoride formation compared with JLF-1 or NIFS-HEAT-2, however weight loss due to WF6 formation was induced, and much Cr2O3 was formed. [ABSTRACT FROM AUTHOR]

[Display omitted] • Micron-scale TiNi matrix with inside nano-scale VCr precipitates became amorphous while micron-scale VCr matrix including nano-scale TiNi precipitates exhibited a high structural stability in V 34 Ti 25 Cr 10 Ni 30 Pd 1 alloy after radiation. • The radiation tolerance of TiNi intermetallic compound in V 34 Ti 25 Cr 10 Ni 30 Pd 1 alloy was increased after nanocrystallization upon irradiation. • VCr solid solution phase exhibits a better radiation tolerance than TiNi intermetallic phase. Intermetallic compounds often have a low radiation tolerance due to the low recombination rate of radiation-induced defects. In the present work, we designed a novel multicomponent vanadium-based alloy (MVA), V 34 Ti 25 Cr 10 Ni 30 Pd 1 , containing a micron-scale TiNi matrix phase with VCr nanoprecipitates (NPs) and a micron-scale VCr matrix phase with TiNi NPs. The MVA was irradiated with 6 MeV Ti 3 + ions with a radiation dose of 5 × 10 15 i o n s / c m 2 at room temperature. Results indicated that the micron-scale intermetallic TiNi matrix along with VCr NPs inside both became amorphous, while the micron-scale VCr matrix including numerous intermetallic TiNi NPs both exhibited a high structural stability after ion irradiation. The intermetallic TiNi matrix becomes amorphous due to the accumulation of radiation-induced defects, and the intermetallic TiNi NPs in VCr matrix have a high stability of crystallographic structure due to high-density interfaces between NPs and matrix. These results indicate that the phase stability of TiNi intermetallic compound is increased after nanocrystallization. Besides, the discrepancy of two matrix phases including precipitates after ion irradiation, and the underlying mechanisms are discussed in detail in this work. This work gives a guidance for designing new vanadium alloys and intermetallic compounds with enhanced structural stability under ion irradiation. [ABSTRACT FROM AUTHOR]

Abstract: Development of ceramic coatings at inner wall of duct tubing is strongly required for self-cooled lithium/vanadium-alloy (Li/V-alloy) blanket concept to reduce magneto-hydrodynamics (MHD) pressure drop. In our previous study, erbium oxide (Er2O3) coatings fabricated by physical vapor deposition (PVD) method showed high electrical resistance and high compatibility with liquid lithium (Li). In situ formation of Er2O3 coatings was also explored to show oxygen in the V-alloy substrate and Er dissolved in Li could react to form thin Er2O3 layer at the interface. In this study, feasibility of self-healing of Er2O3 coating is explored by observing its in situ formation at cracks on the coatings fabricated by PVD method. V–4Cr–4Ti substrates were oxidized at 973K for 6h and annealed at 973K for 16h to introduce proper amount of oxygen into the substrates. The coatings deposited at room temperature (R.T.) were uniformly flat. On the contrary, the coatings deposited at 873K had peeled off at most areas. After introducing cracks on Er2O3 coatings by heating, the samples were exposed in liquid Li doped with Er at 873K for 100h and afterward at 973K for 100h to heal the cracks. At the large peeled-off areas, no in situ formation was observed, while the small cracks were healed. Peel-off of the coating indicates the intrusion of Li between the coating and the substrate. For improvement of self-healing of Er2O3 coatings, it is essential to progress in the adherence of them to the substrates and control the surface condition of the substrates. [Copyright &y& Elsevier]

Abstract: Using V–4Cr–4Ti as structural material in a V–Li blanket offers the blanket to operate at 350–750°C with high thermal efficiency. This paper investigates ways to strengthen V–4Cr–4Ti and the thermal stability of the strengthening. Results show that the alloy could be greatly strengthened in a proper ageing process, with insignificant loss of ductility or even higher static tensile fracture toughness. Cold rolling is found to be another approach for higher strength. Both strengthening are thermally stable at not more than 500°C. At higher temperature, the alloy in the sate of complete recrystallization shows better thermal stability than the alloy in solid solution state. For high performance of the V–Li blanket with good economic feature, the design of the blanket should consider to use V–4Cr–4Ti in various thermo-mechanical states in terms of the operation temperature. [Copyright &y& Elsevier]

Abstract: The in situ Er2O3 insulating coating is under development for the self-cooled Li/V-alloy type fusion blanket. In this study, the structural stability and self-healing capability of the coating are investigated. Since the cracking in the coating was not observed after exposure when Li was removed with a weak lotion (liquid NH3), the cracking observed in the previous studies is not a practical issue in a real blanket. The re-exposure of the coating in pure Li showed that the coating once formed in Li (Er) is thought to be stable in pure Li. Thus, coating has the possibility to be serviced in a Li environment without an Er supply. By prior exposure to Li (Er) at 873K, the exhaustion of the oxygen storage in V-alloy substrate during exposure at 973K could be delayed effectively. The self-healing capability of the coating was demonstrated by the examination with the re-exposing cracked coating in Li (Er). [Copyright &y& Elsevier]

Abstract: Many oxide and nitride coatings are under consideration worldwide for vanadium (V) structure to mitigate the magnetohydrodynamic (MHD) pumping power losses for the self-cooled lithium (Li) blanket system. To date, researchers have focused on the MHD resistance, compatibility, adherence, thermodynamic, and interface issues, paying little attention to the nuclear concerns. This assessment examines the nuclear-related issues for the leading coatings proposed for the self-healing approach in particular. Specifically, we assessed the impact of the coating additives to Li on the breeding potential of a typical Li/V system. An important outcome of this study has been to identify the coatings that compromise the breeding and rank the remaining candidates based on their nuclear behavior using the ARIES-RS operating conditions. It appears likely that the nuclear requirements will have an important impact on the choice of the coating material and on the maximum coating concentration in Li. Potential solutions that mitigate the effect of coatings and compensate for the breeding losses are discussed in the paper. However, these solutions introduce additional requirements and concerns that must be considered in future studies. [Copyright &y& Elsevier]

Abstract: Neutronics calculations were carried out to characterize self-cooled Li and Flibe blankets with vanadium alloy (V–4Cr–4Ti) structure, as advanced options for the Force Free Helical Reactor (FFHR-2). In the present paper, the parameter dependence of Tritium Breeding Ratio (TBR) and neutron shield performance was examined for a Li blanket (FFHR-LV) and a Flibe blanket (FFHR-FV). The TBR was independent from the Flibe composition between 30 and 45mol% of BeF2 in the case of the FFHR-FV. Due to the increase in the temperature range of the coolant Flibe as a result of the change of the structural materials from JLF-1 (low activation ferritic steel) to V–4Cr–4Ti, the fraction of BeF2 can be changed from 43 to 31mol% without changing the margin to coagulation of Flibe. The compositional change of Flibe resulted in a significant decrease in its viscosity. The decrease of TBR by the effects of MHD insulator coating with Er was negligibly small with the expected doping level of Er in the FFHR-LV. The increase in the B4C in the shielding material resulted in the decrease in the fast neutron flux at the superconducting magnet by 60–70% in both FFHR-LV and FFHR-FV. Optimum neutronics performance for the two blanket systems was derived by the calculation. [Copyright &y& Elsevier]

Abstract: The in situ Er2O3 insulating coating is under development for self-cooled Li/V-alloy type fusion blankets. A previous study showed formation and long-term stability of the coating by exposing V–4Cr–4Ti substrate to liquid Li doped with Er. In this study, the mechanisms for nucleation and growth of the coating are investigated. By oxidation and annealing, oxygen was charged into the substrate to form a Ti–O net phase as oxygen source to the coating. The surface layer formed on V–4Cr–4Ti consists of two sub-layers, an insulating Er2O3 coating and an intermediate layer of mixed ErN and Er–V–O. The measured growth rate shows logarithmic kinetics with high exponent (n ≈3 or 4) at 873 and 923K, suggesting that the rate of growth to Er2O3 coating should be very low. The growth was accelerated suddenly at 973K resulting in a low exponent (n ≈2) that almost obeys a parabolic law. A phenomenological model was proposed to describe the mechanism of coating, which showed the growth of the Er2O3 coating is controlled by the diffusion of oxygen and delivery of erbium to interface between V-alloys substrate and liquid Li. [Copyright &y& Elsevier]

Abstract: Helium irradiation experiments of V–4Cr–4Ti alloy with various surface treatments were conducted in an electron cyclotron resonance (ECR) ion irradiation apparatus. After helium ion irradiation at room temperature, the helium thermal desorption and retention properties were examined by thermal desorption spectroscopy (TDS). Ion energy of helium beam was 5keV. Three groups of desorption peaks appeared at around 500, 850 and 1200K in the TDS spectrum. After helium ion irradiation at ion fluence of 1×1021 He/m2, the retained helium desorbed mainly at around 1200K and all of the implanted helium atoms were retained. With increasing fluence up to 5×1021 He/m2, the amount of helium desorbed at 500K increased. For the polished samples with annealing at various temperatures, the desorption peak observed at around 500K shifted to higher temperature region. Smallest retained amount of helium was observed in the V-alloy with polishing followed by annealing at 1373K. [Copyright &y& Elsevier]

Abstract: Ferritic steel, vanadium alloy and SiC/SiC composite are candidate low activation materials for blanket components and first walls in fusion demonstration reactors. Several issues on these materials as the first wall have been investigated so far. Amount of deuterium retained in mechanically polished ferritic steel, F82H, after deuterium ion irradiation, was observed to be several times smaller than that of stainless steel, 316L SS. Physical sputtering yield of the ferritic steel due to deuterium ion was comparable to that of 316L SS. These results suggest that the property of the ferritic steel as the first wall material is superior to that of 316L SS, with respect to fuel hydrogen retention and in-vessel tritium inventory. Since first walls of blanket modules are exposed to both fuel hydrogen and helium, the helium is also trapped in the walls. Helium retention of V–4Cr–4Ti was investigated using helium ion irradiation apparatus. The amount of helium retained was comparable to those of other plasma facing materials. One of the major concerns in use of SiC/SiC composite for blanket is permeation of helium gas coolant into fusion plasma. Helium gas permeability of the SiC/SiC composite after heat cycles was measured using a vacuum device consisting two chambers. The increase in the permeability was not observed when the heating rate was suitably adjusted. Therefore, the blanket module may be made using only SiC/SiC composite if a vacuum pumping for the inside of blanket module is attached. [Copyright &y& Elsevier]

Abstract: MHD pressure drop in flowing liquid metal for a tokamak with high magnetic fields is a key concern regarding the development of lithium self-cooled test module for ITER and lithium breeding blanket for DEMO-type projects. MHD losses on liquid metal pumping can be most efficiently reduced by applying of an electrically insulating coating to the inner surface of the channels. The developed method to evaluate the electrically insulating properties of a coating placed on the internal surface of a vanadium test channel is based on the pressure drop measurement in a liquid lithium forced circulation system with the test section in magnetic field. The tests were conducted on channels of V–4Ti–4Cr alloy with and without on insulating coating based on the aluminum nitride technique (AlN) . The dimensions of the test section were 6mm×20mm×320mm. The tests involved lithium flow velocities up to 3m/s, temperatures up to 400°C and a uniform transverse magnetic field up to 1.6T. The measured hydraulic resistances have shown five times reduction for the coated wall channel in comparison with the conducting wall channel. [Copyright &y& Elsevier]

Abstract: An electrically insulating coating is under development to mitigate magneto-hydrodynamic (MHD) pressure drop in self-cooled Li/V-alloy blanket. A previous study showed the feasibility of in situ formation of an Er2O3 coating by exposing V–4Cr–4Ti to liquid Li doped with Er at high temperature. In this paper, the characteristics and the long-term stability of such a coating were investigated. A V–4Cr–4Ti substrate was oxidized in flowing Ar at 973K, annealed in vacuum at 973K and finally exposed to liquid Li doped with Er from 773 to 973K. The oxygen charged and homogenized at the surface of the substrate that acted as a source for coating formation. The surface layer formed on V–4Cr–4Ti consists of double sub-layers, namely an insulating coating of Er2O3 and an intermediate layer of mixed ErN and V-compounds. The Er2O3 coating was found stable up to 750h of exposure time in liquid Li at 873K with its saturated thickness of ∼0.1μm. The coating was stable up to 300h at 973K with its thickness increased to ∼1.2μm. Cracks were observed in coatings formed at 923 and 973K. The mechanism of cracking is under investigation. [Copyright &y& Elsevier]

Helium irradiation experiments of V–4Ti alloy were conducted in an ECR ion irradiation apparatus by using helium ions with energy of 5keV. The ion fluence was in the range from 1×1017He/cm2 to 8×1017He/cm2. After the helium ion irradiation, the helium retention was examined by using a technique of thermal desorption spectroscopy (TDS). After the irradiation, the blisters with a size of about 0.1μm were observed at the surface, and the blister density increased with the ion fluence. Two desorption peaks were observed at approximately 500 and 1200K in the thermal desorption spectrum. When the ion fluence was low, the retained helium desorbed mainly at the higher temperature regime. As increase of the ion fluence, the desorption at the lower temperature peak increased and the retained amount of helium saturated. The saturated amount was approximately 2.5×1017He/cm2. This value was comparable with those of the other plasma facing materials such as graphite. [Copyright &y& Elsevier]

Corrosion behaviour of Aluminum nitride (AlN) as a candidate material for insulating coating for V/Li blanket was investigated by corrosion experiments in liquid lithium (Li) up to 1073 K for 1000 h. High purity AlN samples decreased their weights after the sintering test in Li in contact with vanadium alloy over 973 K, while those in Li not in contact with the vanadium alloy survived up to 1073 K. Nitrogen dissolution from AlN into the liquid Li and absorption of dissolved nitrogen by vanadium alloy are considered to be a corrosion mechanism. Small decreases of electrical resistance were observed after the sintering tests over 873 K because of the conductive corrosion layer on the surface caused by the nitrogen dissolution. In the cases of low purity AlN samples sintered over 973 K, large weight decreases were explained by fragile grain boundary caused by oxygen dissolution. The oxygen dissolution may also results in the resistivity decrease even at 723 K. Therefore, decrease of oxygen impurity in AlN and addition of nitrogen in liquid Li are considered to give a possible solution to the Li/AlN corrosion problem in the Li/AlN/V alloy blanket system. [Copyright &y& Elsevier]

Microstructure and irradiation hardening of V-4Cr-4Ti alloys with different yttrium (Y) contents were studied by self-ion irradiation at 550 °C via TEM and nano-indentation test technology. The peak damage of the V-4Cr-4Ti-xY alloy (x = 0, 0.1, 0.5, and 1, wt.%) irradiated by 2.5 MeV self-ion (V2+) is 8 dpa. Dense dislocation loops were observed in all vanadium alloy samples after irradiation. With the increase of Y content, both average size and number density of dislocation loops using g = <110> near the pole [001] decreased, while the irradiation hardness increment first decreased and then increased. In order to better reduce the irradiation hardening, it is considered that the addition of 0.1 wt.% Y in V-4Cr-4Ti alloy is reasonable. [ABSTRACT FROM AUTHOR]