Microstructure, oxidation behaviour and thermal shock resistance of self-passivating W-Cr-Y-Zr alloys

Sal, E.; Garcı́a-Rosales, C.; Schlueter, K.; Hunger, K.; Gago, Mauricio; Wirtz, M.; Calvo, A.; Andueza, Iñigo; Neu, R.; Pintsuk, G.

doi:10.1016/j.nme.2020.100770

Cited by 7 publications

(3 citation statements)

References 22 publications

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“…However, the lower density of the supplied samples results in a slightly lower strength compared to samples with the usual density. The HTed, dense W-10Cr-0.5Y alloy has withstand 1000 thermal cycle loads of 0.38 GW m −2 power density and 1 ms duration with the appearance of a crack network while the addition of 0.5 wt.% Zr to this alloy results in no damage after the same loading [52].…”

Section: Materials Irradiation and Testing Methodologymentioning

confidence: 99%

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Terentyev

Jenuš²,

Broco³

et al. 2022

Nucl. Fusion

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Development of refractory metals for application as plasma-facing armour material remains among priorities of fusion research programmes in Europe, China and Japan. Improving the resistance to high temperature recrystallization, enhancing material strength to sustain thermal fatigue cracking and tolerance to neutron irradiation are the key indicators used for the down selection of materials and manufacturing processes to be applied to deliver engineering materials. In this work we investigate the effect of neutron irradiation on mechanical properties and microstructure of several tungsten grades recently developed. Neutron irradiation campaign is arranged for screening purposes and therefore is limited to the fluence relevant for the ITER plasma facing components. At the same time, the neutron exposure covers a large span of irradiation temperatures from 600 up to 1000 °C. Four different grades are included in the study, namely: fine-grain tungsten strengthened by W-carbide (W-4wt.% W2C), fine-grain tungsten strengthened by Zr carbides (W-0.5% ZrC), W alloyed with 10 at.% chromium and 0.5 at.% yttrium (W-10Cr-0.5Y) and technologically pure W plate manufactured according to the ITER specification by Plansee (Austria). The strengthening by W2C and ZrC particles leads to an enhanced strength, moreover, the W-0.5ZrC material exhibits reduced DBTT (compared to ITER specification grade) and is available in the form of thick plate (i.e. high up-scaling potential). The W-10Cr-0.5Y grade is included as the material offering the self-passivation protection against the high temperature oxidation.

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Section: Materials Irradiation and Testing Methodologymentioning

confidence: 99%

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Terentyev

Jenuš²,

Broco³

et al. 2022

Nucl. Fusion

Self Cite

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“…The first step in the production is the mechanical alloying of the source W, Cr and Y powders. This step is treated as virtually inevitable and used by all research groups working on self-passivating alloys worldwide [11,22,[24][25][26][27][28]. The following step is the sintering of the alloyed powder.…”

Section: Oxidation Performancementioning

confidence: 99%

Advanced Self-Passivating Alloys for an Application under Extreme Conditions

Litnovsky

Klein

Tan

et al. 2021

Metals

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Self-passivating Metal Alloys with Reduced Thermo-oxidation (SMART) are under development for the primary application as plasma-facing materials for the first wall in a fusion DEMOnstration power plant (DEMO). SMART materials must combine suppressed oxidation in case of an accident and an acceptable plasma performance during the regular operation of the future power plant. Modern SMART materials contain chromium as a passivating element, yttrium as an active element and a tungsten base matrix. An overview of the research and development program on SMART materials is presented and all major areas of the structured R&D are explained. Attaining desired performance under accident and regular plasma conditions are vital elements of an R&D program addressing the viability of the entire concept. An impressive more than 104-fold suppression of oxidation, accompanied with more than 40-fold suppression of sublimation of tungsten oxide, was attained during an experimentally reproduced accident event with a duration of 10 days. The sputtering resistance under DEMO-relevant plasma conditions of SMART materials and pure tungsten was identical for conditions corresponding to nearly 20 days of continuous DEMO operation. Fundamental understanding of physics processes undergone in the SMART material is gained via fundamental studies comprising dedicated modeling and experiments. The important role of yttrium, stabilizing the SMART alloy microstructure and improving self-passivating behavior, is under investigation. Activities toward industrial up-scale have begun, comprising the first mechanical alloying with an industrial partner and the sintering of a bulk SMART alloy sample with dimensions of 100 mm × 100 mm × 7 mm using an industrial facility. These achievements open the way to further expansion of the SMART technology toward its application in fusion and potentially in other renewable energy sources such as concentrated solar power stations.

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“…Elisa et al prepared a W-Cr-Y-Zr alloy through mechanical alloying and HIP. Oxidation tests conducted at 1000 • C showed that the alloy containing Zr exhibited significantly greater oxidation resistance compared to pure W. When subjected to thermal shock loading simulating 1000 ELM-like pulses at the divertor, the heat-treated Zrcontaining alloy did not exhibit any damage [19]. We prepared a W-Si alloy by mechanical alloying (MA) and spark plasma sintering (SPS).…”

Section: Introductionmentioning

confidence: 99%

Microstructures and Antioxidation of W Self-Passivating Alloys: Synergistic Effect of Yttrium and Milling Time

Chen,

Xue,

Yin

et al. 2024

Metals

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Tungsten and its alloys are widely recognized as key components in high-temperature environments. In this study, self-passivating W-Si-xY alloys with varying Y content were prepared using mechanical alloying (MA) and spark plasma sintering (SPS). The synergistic effects of Y content and milling time on the microstructures and oxidation resistance of the alloys were revealed. This study found that the oxidation resistance of the alloys increased as the Y content increased. However, the effect of milling time on oxidation resistance was complex. For W-Si-xY alloys with low Y content (0Y and 2Y), the oxidation resistance decreased with increasing milling time. In contrast, for W-Si-xY alloys with high Y content (4Y and 6Y), the oxidation resistance increased with increasing milling time. This enhanced oxidation resistance is due to the microstructural changes in the protective composite layer, including the size and distribution of W5Si3, Y2Si2O7 aggregates, and W-Y-O melt. The thickness of the oxide layer on the W-Si-6Y alloy after being oxidized at 1000 °C for 2 h was only 70.7 μm, demonstrating its superior oxidation resistance.

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Microstructure, oxidation behaviour and thermal shock resistance of self-passivating W-Cr-Y-Zr alloys

Cited by 7 publications

References 22 publications

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Advanced Self-Passivating Alloys for an Application under Extreme Conditions

Microstructures and Antioxidation of W Self-Passivating Alloys: Synergistic Effect of Yttrium and Milling Time

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