High temperature microstructural stability of self-passivating W-Cr-Y alloys for blanket first wall application

Sal, E.; Garcı́a-Rosales, C.; Iturriza, I.; Andueza, Iñigo; Burgos, N.

doi:10.1016/j.fusengdes.2019.02.136

Cited by 17 publications

(5 citation statements)

References 11 publications

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“…After ageing at 1000 • C for 50 and 100 h, the microstructure exhibits a phase decomposition with significant grain refinement (figure 6(a)). After 50 h, the Cr-rich phase presents a lamellar shape while after 100 h, a vermicular shape typical for spinodal decomposition has been developed [49]. Based on these results, a complete phase decomposition with vermicular microstructure and large grain refinement is predicted after the irradiation test of W-10Cr-0.5Y samples annealed at 1000 • C lasting for more than 50 days.…”

Section: W-wcmentioning

confidence: 75%

“…The forth material is an alloy of composition W-10Cr-0.5Y in wt.% prepared from elemental powders of W (99.95%, 20 μm), Cr (99.95%, 50-70 μm) and Y (99.9%, 20-30 μm) which are mechanically alloyed (MA) under Ar atmosphere in a planetary ball mill using WC grinding jars and balls [49]. After MA, a single (αW, Cr) nanocrystalline supersaturated phase is achieved.…”

Section: Materials Irradiation and Testing Methodologymentioning

confidence: 99%

“…Besides, elemental Y is expected to combine with practically all residual oxygen introduced during the MA process, forming Y 2 O 3 and cleaning the grain boundaries (GB) from oxygen. The alloyed powder is introduced in metallic capsules, which are evacuated, degassed, sealed and HIPed at 1250 • C for 2 h at 150 MPa [49]. After HIPing, a homogeneous, ultrafine grained microstructure is obtained with an average grain size of the equiaxed matrix below 100 nm.…”

Section: Materials Irradiation and Testing Methodologymentioning

confidence: 99%

“…Mechanical characterization [23] from RT to 1000 • C has Neutron irradiation tests were performed at irradiation temperatures of 600, 800 and 1000 • C for 50 days. Since the W-10Cr-0.5Y alloy after HT presents a metastable microstructure, samples were exposed in a previous work [49] to a temperature of 650 • C for 3000 h to study the thermal stability of the microstructure under operational conditions. After the ageing test, there were no significant changes, as the microstructure of the material as well as its density and hardness remain unchanged (figure 6(c)).…”

Section: W-wcmentioning

confidence: 99%

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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: W-wcmentioning

confidence: 75%

Section: Materials Irradiation and Testing Methodologymentioning

confidence: 99%

Section: Materials Irradiation and Testing Methodologymentioning

confidence: 99%

Section: W-wcmentioning

confidence: 99%

See 2 more Smart Citations

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Terentyev

Jenuš²,

Broco³

et al. 2022

Nucl. Fusion

Self Cite

View full text Add to dashboard Cite

show abstract

“…Main achievements in the area of W-Cr-Y are the inclusion of yttrium into the ternary systems [69,83] and the development of two main production routes: FAST [85] by the group at Forschungszentrum Jülich GmbH and hot isostatic pressing (HIP) at CEIT (Spain). [99,100] The detailed understanding of the underlying oxidation and sublimation mechanisms is also important. [85,86] A new system under consideration is currently W-Cr-Zr.…”

Section: W Alloysmentioning

confidence: 99%

Fusion Materials Development at Forschungszentrum Jülich

Coenen

2020

Adv Eng Mater

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Material issues pose a significant challenge for the design of future fusion reactors. From a historic point of view, the material mix used for the first wall of a fusion reactor has continually evolved, from original steel vessels to carbon and other low-Z materials such as beryllium to tungsten as the primary candidate for a reactor's first wall armor and divertor material. For materials considered for fusion applications, a highly integrated approach is necessary. Resilience against neutron damage, good power exhaust, and oxidation resistance during accidental air ingress are design relevant issues while deciding on new materials or improving upon baseline materials. Neutron-induced effects, e.g., transmutation adding to embrittlement, retention, and changes to thermomechanical properties, are crucial to material performance. In this contribution, the recent progress (2013-2019) in fusion materials development for current and future fusion devices, at Forschungszentrum Jülich GmbH, with activities focussing on advanced materials and their characterization is given. It is a continuation and extension of the work given by Coenen et al.

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Microstructural and mechanical characterization of self-passivating W-Eurofer joints processed by brazing technique

Prado

Sal

Sánchez

et al. 2021

Fusion Engineering and Design

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High temperature microstructural stability of self-passivating W-Cr-Y alloys for blanket first wall application

Cited by 17 publications

References 11 publications

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Development of irradiation tolerant tungsten alloys for high temperature nuclear applications

Fusion Materials Development at Forschungszentrum Jülich

Microstructural and mechanical characterization of self-passivating W-Eurofer joints processed by brazing technique

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