Tensile properties of baseline and advanced tungsten grades for fusion applications

Yin, Chuntao; Terentyev, D.; Pardoen, Thomas; Bakaeva, A.; Petrov, Roumen; Antusch, S.; Rieth, M.; Vilémová, Monika; Matějíček, Jiří; Zhang, Tao

doi:10.1016/j.ijrmhm.2018.04.003

Cited by 64 publications

(26 citation statements)

References 29 publications

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“…Based on the review of the available open literature data, one can single out three providers of tungsten products that have been fabricated according to the ITER specification, with dimensions suitable for the application in ITER. These providers include: (i) Plansee (Austria), see, e.g., [3,[10][11][12][13][14][15][16][17]; (ii) Advanced Technology and Materials AT&M (China), see, e.g., [18][19][20][21][22][23][24][25], and A.L.M.T. (Japan), see, e.g., [13,[26][27][28][29][30][31].…”

Section: Introductionmentioning

confidence: 99%

Neutron irradiation effects on mechanical properties of ITER specification tungsten

et al. 2021

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In this contribution, we present the results of recent neutron irradiation campaign performed in the material test reactor BR2 (Belgium) on pure tungsten. We have applied various irradiation conditions and sample geometry to assess the effect of neutron irradiation on hardness, bending, tensile and fracture mechanical properties. The investigated material is a commercially pure tungsten plate fabricated according to the international thermonuclear experimental reactor (ITER) specification for the application in the divertor plasma-facing components. The neutron irradiation covers a large span of temperatures and damage doses, ranging from 600 to 1200 °C and 0.1-1 dpa. The obtained mechanical properties were analyzed to deduce the shift of the ductile to brittle transition temperature (DBTT) applying bending, tensile and fracture toughness-testing procedures. Then, a correlation of the fracture toughness with the change of the hardness was established. The obtained results are compared with the already published results on another ITER specification grade produced in the form of a rod. The presented and discussed results show that the performance of the compared grades in terms of the irradiation-induced embrittlement is similar, and that the irradiation in the high-temperature region (600-800 °C) causes a considerable DBTT shift already at 0.2-0.5 dpa.

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Section: Introductionmentioning

confidence: 99%

Neutron irradiation effects on mechanical properties of ITER specification tungsten

et al. 2021

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“…the W-2%Y 2 O 3 case). However, W-1%TiC consistently shows better crack resistance than the other two counterparts, which may be related to its different phase boundary structures between the TiC ceramic particle and the W matrix [48], not captured by our simple model. Furthermore, the observed particle-matrix interface decohesion for the two composites may be relevant but was not considered.…”

Section: Surface Crack Formationmentioning

confidence: 80%

Fracture behavior of tungsten-based composites exposed to steady-state/transient hydrogen plasma

et al. 2020

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“…Figure 3 shows the thermally loaded surfaces of rolled W-0.5ZrC after exposure to single shot with a pulse length of 5 ms using an electron beam. No cracks or melting were observed on the samples when tested at an absorbed power The tensile stress of various W materials [51]. When the APD increased to 1.1 GW/m 2 (Figure 3c), both cracks and melting were found on the sample surface, indicating a crack threshold of 0.88-1.1 GW/m 2 .…”

Section: Carbide Dispersion-strengthened Tungsten-based Materialsmentioning

confidence: 98%

“…The W-3%Re and K-doped W-3%Re show ductility at 100°C, which is comparable with W-0.5ZrC plate, while their tensile strength values (<1000 MPa) are lower than that of W-0.5ZrC plate (1058 MPa) [50]. As compared with Plansee ITER specification W (IGP), AT&M ITER specification W (CEFTR), W-1wt%TiC (W1TiC), W-2wt%Y 2 O 3 (W2YO) from Karlsruhe Institute of Technology at Germany, and fine-grained W (FG) from the Institute of Plasma Physics at Czech Republic, W-0.5ZrC exhibits the highest tensile strength and lower DBTT [51], as shown in Figure 2.…”

Section: Carbide Dispersion-strengthened Tungsten-based Materialsmentioning

confidence: 99%

The Tungsten-Based Plasma-Facing Materials

Zhang¹,

Xie²,

Liu³

et al. 2020

Fusion Energy

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The plasma-facing materials in fusion reactors will face very extreme servicing condition such as high temperatures, high thermal loads, extreme irradiation conditions induced by high-energy neutron, and high fluences of high-flux and low-energy plasma. Tungsten is considered as the most promising material for plasma-facing components (PFCs) in the magnetic confinement fusion devices, due to its high melting temperature, high thermal conductivity, low swelling, low tritium retention, and low sputtering yield. However, some important shortcomings such as the irradiation brittleness and high ductility-brittle transition temperature of pure tungsten limit its application. Focusing on this issue, various W alloys with enhanced performance have been developed. Among them, nanoparticle dispersion strengthening such as oxide particle dispersion-strengthened (ODS-W) and carbide particle dispersion-strengthened (CDS-W) tungsten alloys and W fiber-reinforced W f /W composites are promising. This chapter mainly reviews the preparation, microstructure, properties, regulation, and service performance evaluation of ODS-W, CDS-W, and W f /W materials, as well as future possible development is proposed.

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Tensile properties of baseline and advanced tungsten grades for fusion applications

Cited by 64 publications

References 29 publications

Neutron irradiation effects on mechanical properties of ITER specification tungsten

Neutron irradiation effects on mechanical properties of ITER specification tungsten

Fracture behavior of tungsten-based composites exposed to steady-state/transient hydrogen plasma

The Tungsten-Based Plasma-Facing Materials

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