Fusion materials research started in the early 1970s following the observation of the degradation of irradiated materials used in the first commercial fission reactors. The technological challenges of fusion energy are intimately linked with the availability of suitable materials capable of reliably withstanding the extremely severe operational conditions of fusion reactors. Although fission and fusion materials exhibit common features, fusion materials research is broader. The harder mono-energetic spectrum associated with the deuterium–tritium fusion neutrons (14.1 MeV compared to <2 MeV on average for fission neutrons) releases significant amounts of hydrogen and helium as transmutation products that might lead to a (at present undetermined) degradation of structural materials after a few years of operation. Overcoming the historical lack of a fusion-relevant neutron source for materials testing is an essential pending step in fusion roadmaps. Structural materials development, together with research on functional materials capable of sustaining unprecedented power densities during plasma operation in a fusion reactor, have been the subject of decades of worldwide research efforts underpinning the present maturity of the fusion materials research programme. For achieving proper safety and efficiency of future fusion power plants, low-activation materials able to withstand the extreme fusion conditions are needed. Here, the irradiation physics at play and fusion materials research is reviewed.

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Materials research for fusion

On the eve of component procurement, this paper discusses the present physics basis for the first ITER tungsten (W) divertor, beginning with a reminder of the key elements defining the overall design, and outlining relevant aspects of the Research Plan accompanying the new “staged approach” to ITER nuclear operations which fixes the overall divertor lifetime constraint. The principal focus is on the main design driver, steady state power fluxes in the DT phases, obtained from simulations using the 2-D SOLPS-4.3 and SOLPS-ITER plasma boundary codes, assuming the use of the low Z seeding impurities nitrogen (N) and neon (Ne). A new perspective on the simulation database is adopted, concentrating purely on the divertor physics aspects rather than on the core-edge integration, which has been studied extensively in the course of the divertor design evolution and is published elsewhere. Emphasis is placed on factors which may increase the peak steady state loads: divertor target shaping for component misalignment protection, the influence of fluid drifts, and the consequences of narrow scrape-off layer heat flux channels. All tend to push the divertor into an operating space at higher sub-divertor neutral pressure in order to remain at power flux densities acceptable for the target material. However, a revised criterion for the maximum tolerable loads based on avoidance of W recrystallization, sets an upper limit potentially ∼50% higher than the previously accepted value of ∼10 MW m−2, a consequence both of the choice of material and the finalized component design. Although the simulation database is currently restricted to the 2-D toroidally symmetric situation, considerable progress is now also being made using the EMC3-Eirene 3-D code suite for the assessment of power loading in the presence of magnetic perturbations for ELM control. Some new results for low input power corresponding to the early H-mode operation phases are reported, showing that even if realistic plasma screening is taken into account, significant asymmetric divertor heat fluxes may arise far from the unperturbed strike point. The issue of tolerable limits for transient heat pulses is an open and key question. A new scaling for ELM power deposition has shown that whilst there may be more latitude for operation at higher current without ELM control, the ultimate limit is likely to be set more by material fatigue under large numbers of sub-threshold melting events.

Physics basis for the first ITER tungsten divertor

Plasma-facing materials and components in a fusion reactor are the interface between the 
plasma and the material part. The operational conditions in this environment are probably 
the most challenging parameters for any material: high power loads and large particle and 
neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak 
or stellarator reactor, given the proven geometries and technological solutions, requires an 
improvement of the thermo-mechanical capabilities of currently available materials. In its 
first part this article describes the requirements and needs for new, advanced materials for the 
plasma-facing components. Starting points are capabilities and limitations of tungsten-based 
alloys and structurally stabilized materials. Furthermore, material requirements from the 
fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, 
defining directions for the material development. Finally, safety requirements for a fusion 
reactor with its specific accident scenarios and their potential environmental impact lead to 
the definition of inherently passive materials, avoiding release of radioactive material through 
intrinsic material properties. The second part of this article demonstrates current material 
development lines answering the fusion-specific requirements for high heat flux materials. 
New composite materials, in particular fiber-reinforced and laminated structures, as well 
as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical 
operation space towards regions of extreme steady-state and transient loads. Self-passivating

/pdf/development-of-advanced-high-heat-flux-and-plasma-facing-4cdcf31umx.pdf

Development of advanced high heat flux and plasma-facing materials

Since the ITER divertor design includes tungsten monoblocks in the vertical target where heat loads are maximal, the design to protect leading edges as well as technology R&D for high performance armor-heat sink joint were necessary to be implemented. In the R&D, the availability of the technology was demonstrated by high heat flux test of tungsten monoblock components. Not systematically but frequently macro-cracks appeared at the middle of monoblocks after 20 MW/m2 loading. The initiation of such macro-cracks was considered to be due to cyclic exposure to high temperature, ∼2000 °C, where creep, recrystallization and low cycle fatigue were concerned. To understand correlation between the macro-crack appearance and mechanical properties and possible update of acceptance criteria in the material specification, an activity to characterize the tungsten monoblocks was launched.

Use of Tungsten Material for the ITER Divertor

https://publikationen.bibliothek.kit.edu/1000058757/15098709

Overview of the design approach and prioritization of R&D activities towards an EU DEMO

Interpretation of the deep cracking phenomenon of tungsten monoblock targets observed in high-heat-flux fatigue tests at 20 MW/m2

https://pure.mpg.de/pubman/item/item_2070919_5/component/file_2389889/Li_Low.pdf

Low cycle fatigue behavior of ITER-like divertor target under DEMO-relevant operation conditions

Fracture mechanical analysis of tungsten armor failure of a water-cooled divertor target

Influence of heat flux loading patterns on the surface cracking features of tungsten armor under ELM-like thermal shocks

In this work, the cracking behavior of tungsten under edge-localized mode (ELM)-like thermal shock loads was investigated on the basis of a rigorous computational fracture mechanical analysis combined with the finite element method. Typical transient thermal shock loads of ELM conditions were considered with a relevant range of power density and base temperature for a loading duration of 1 ms. Crack initiation and progressive growth were predicted using the extended finite element method and the J -integral was calculated for the assumed precrack by means of the virtual crack extension method. For a power density of 1 GW/m 2 and higher, a crack is preferably initiated near the edge of the loading area and is then followed by a gradual horizontal kinking, parallel to the loading surface. The crack formation is predicted for the power density of 0.6 GW/m 2 and above, and when the base temperature is higher than 600 °C, almost no cracks is predicted. The numerically predicted cracking behavior agrees in general with the experimental observations.

Muyuan Li

Papers

Interpretation of the deep cracking phenomenon of tungsten monoblock targets observed in high-heat-flux fatigue tests at 20 MW/m2

Low cycle fatigue behavior of ITER-like divertor target under DEMO-relevant operation conditions

Fracture mechanical analysis of tungsten armor failure of a water-cooled divertor target

Influence of heat flux loading patterns on the surface cracking features of tungsten armor under ELM-like thermal shocks

Cracking behavior of tungsten armor under ELM-like thermal shock loads: A computational study