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Recent progress in research on tungsten materials for nuclear fusion applications in Europe

Michael Rieth, +70 more
- 01 Jan 2013 - 
- Vol. 432, Iss: 1, pp 482-500
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TLDR
In this article, the progress of work within the EFDA long-term fusion materials program in the area of tungsten alloys is reviewed, with a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments.
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This article is published in Journal of Nuclear Materials.The article was published on 2013-01-01 and is currently open access. It has received 599 citations till now.

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Nuclear fusion

TL;DR: The advantages of nuclear fusion as an energy source and research progress in this area are summarized in this article, where the current state of the art is described, including the Compact Ignition Tokamak (CIT), International Thermonuclear Experimental Reactor (ITER), and a US design called TIBER II.
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Designing Radiation Resistance in Materials for Fusion Energy

TL;DR: In this article, three fundamental options for designing radiation resistance are outlined: Utilize matrix phases with inherent radiation tolerance, select materials in which vacancies are immobile at the design operating temperatures, or engineer materials with high sink densities for point defect recombination.
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Primary radiation damage: A review of current understanding and models

TL;DR: In this article, the authors consider the extensive experimental and computer simulation studies that have been performed over the past several decades on what the nature of the primary damage is, and provide alternatives to the current international standard for quantifying this energetic particle damage, the Norgett-Robinson-Torrens displacements per atom (NRT-dpa) model for metals.
Journal ArticleDOI

Materials research for fusion

TL;DR: Fusion materials research started in the early 1970s following the observation of the degradation of irradiated materials used in the first commercial fission reactors as mentioned in this paper, and has been the subject of decades of worldwide research efforts underpinning the present maturity of the fusion materials research program.
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Radiation damage in nanostructured materials

TL;DR: In this paper, the authors summarized and analyzed the current understandings on the influence of various types of internal defect sinks on reduction of radiation damage in primarily nanostructured metallic materials, and partially on nanoceramic materials.
References
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Journal ArticleDOI

Simulation of irradiation induced deuterium trapping in tungsten

TL;DR: In this article, the evolution of defects, hydrogen, and impurities in tungsten during and after deuterium irradiation is simulated by solving rate theory equations, and the results are in excellent agreement with irradiation experiments, and they conclude that formed impurity self-interstitial atom complexes could be the nucleation site for formation of large interstitial type dislocation loops observed experimentally.
Journal ArticleDOI

Oxidation behaviour of silicon-free tungsten alloys for use as the first wall material

TL;DR: The use of self-passivating tungsten alloys as armour material of the first wall of a fusion power reactor may be advantageous concerning safety issues as discussed by the authors, however, the formation of brittle tengsten silicides may be disadvantageous for the powder metallurgical production of bulk W?Cr?Si alloys if a good workability is needed.
Journal ArticleDOI

Fracture behaviour of polycrystalline tungsten

TL;DR: In this article, the authors studied the influence of the anisotropic microstructure on the fracture toughness and ductile-to-brittle transition (DBT) of polycrystalline tungsten.
Journal ArticleDOI

Damage structure in divertor armor materials exposed to multiple ITER relevant ELM loads

TL;DR: In this paper, the damage threshold and damage mechanisms of divertor armor materials, i.e., CFC and tungsten, were studied under the impact of ITER relevant ELM-like loads.
Journal ArticleDOI

Atomistic simulation of single kinks of screw dislocations in α-Fe

TL;DR: In this paper, the authors studied the structure and the formation and migration energies of single kinks in ½ ǫ 1 ¼ 1 ¾ 1 1 1/1/ǫ screw dislocations in body-centered cubic iron, by performing static calculations using the Ackland-Mendelev empirical potential.
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Q1. What contributions have the authors mentioned in the paper "Recent progress in research on tungsten materials for nuclear fusion applications in europe" ?

In this paper, the authors present a detailed review of the use and properties of tungsten materials for the first wall of a fusion tokamak. 

The main requirements of tungsten materials for structural divertor applications compriseproperties like high thermal conductivity, high temperature strength and stability, high recrystallization temperature, and enough ductility for an operation period of about two years under enormous neutron load. 

It was demonstrated that diffusion bonding at only 700 °C can be successfully performed with a minimal alteration of the microstructure of the base materials and strongly reduced formation of vanadium-carbide. 

The linear oxidation rate of tungsten at 1000 °C is about 1.410-2 mg cm-2 s-1 [43], which in the approximately 1000 m2 DEMO first wall would correspond to a release of half a ton of tungsten oxides per hour. 

Powder injection moulding (PIM) was investigated as a mass fabrication option for the tungstenarmour tiles which had to be joined to the thimbles. 

After the implantation has stopped (region III), the D retention drops and a remnant D flux to the W surface takes place for about 10 min. 

The use of tungsten as first wall (FW) armour of a fusion power reactor represents an importantsafety concern in the event of a loss of coolant accident with simultaneous air ingress into the reactor vessel. 

But even neglecting the irradiation effects (due to the large gaps in the knowledge of properties of these materials), there are still unsolved problems related to the use and properties of tungsten materials. 

Due to the fabrication route, missing mechanical working and/or an increased impurity level could also be an additional explanation for this severe brittleness. 

In helium cooled divertor designs tungsten materials are also considered for structural use (e.g. as pressurized pipes or thimbles). 

The formation of intermetallic compounds in tungsten alloys is just one of the factors responsible for their increased hardening, the other being the conventional solute hardening that gives rise to the embrittlement of the alloys that occurs even in the limit where the concentration of the alloying elements is small. 

It shows that the vacancy formation energy in W-Ta alloys depends sensitively on the lattice site at which a vacancy is formed, whereas in W-V alloys it is almost independent of the location of the vacancy site. 

In the first case, pure metallic powders were mechanically mixed, compacted, and molten to allow for brazing filler materials with homogeneous compositions. 

In a recent development, precursor powders are fabricated under certain solution conditions where the particle growth could be controlled to produce uniformly yttrium doped nano-sized tungsten oxides. 

In W, the production of impurities, such as Re, Ta, and Os, is fairly significant, being of the order of a few thousand to tens of thousands of atomic parts per million (appm) over a typical DEMO-like first-wall 5-year neutron exposure. 

Preliminary results of mechanical characterization of these W-W joints using Ti-Fe fillers gave rise to an average shear strength of 140 ± 8 MPa. 

So far, only rhenium is known to improve the ductility of tungsten by solid solution but its usefor fusion energy applications has been ruled out for various reasons (cost, irradiation embrittlement). 

The oxidation behaviour of the WCr12Ti2.5 alloy is similar to that of the WCr10Si10 material; in this case the oxidation rate is similar to that of thin films of same composition at 600 °C but higher at 800 and 1000 °C [48]. 

Only grain growing and physical sputtering were identified as the surface modification processes of the pure hydrogen loaded materials (Fig. 17). 

The results show the possibility of designing alloys where vacancies form within a desired range of temperature, suggesting the possibility of developing alloys with improved stability under irradiation. 

the fracture resistance may increase with crack propagation, which implies that it is not always possible to characterize the material‘s toughness with one single value such as plane strain fracture toughness KIC or critical energy release rate GIC. 

In what follows, the results, conclusions, and outlooks are summarized for each of theW&WALLOYS programme‘s main subtopics, which are (1) fabrication, (2) structural W materials, (3) W armour materials, and (4) materials science and modelling. 

Further investigations into the effect of grain boundary crystallography and chemistry are currently on-going, but it appears that tantalum has no beneficial effects, and may even have detrimental effects, on the fracture properties of tungsten.