Behavior of tungsten under irradiation and plasma interaction

High entropy alloys are multicomponent alloys, which consist of five or more elements in equiatomic or nearly equiatomic concentrations. These materials are hypothesized to show significantly decreased self-diffusivities. For the first time, diffusion of all constituent elements in equiatomic CoCrFeNi and CoCrFeMnNi single crystals and additionally solute diffusion of Mn in the quaternary alloy is investigated using the radiotracer technique, thereby the tracer diffusion coefficients of 57Co, 51Cr, 59Fe, 54Mn, and 63Ni are determined at a temperature of 1373 K. The components are characterized by significantly different diffusion rates, with Mn being the fastest element and Ni and Co being the slowest ones. Furthermore, solute diffusion of Cu in the CoCrFeNi single crystal is investigated in the temperature range of 973–1173 K using the 64Cu isotope. In the quaternary alloy, Cu is found to be a fast diffuser at the moderate temperatures below 1273 K and its diffusion rate follows the Arrhenius law with an activation enthalpy of about 149 kJ/mol.

/pdf/tracer-diffusion-in-single-crystalline-cocrfeni-and-ddm73g349g.pdf

Tracer diffusion in single crystalline CoCrFeNi and CoCrFeMnNi high entropy alloys

This paper presents the roadmap of the main materials to be used for ITER and DEMO class reactors as well as an overview of the most relevant innovations that have been made in recent years. The main idea in the EUROfusion development program for the FW (first wall) is the use of low-activation materials. Thus far, several candidates have been proposed: RAFM and ODS steels, SiC/SiC ceramic composites and vanadium alloys. In turn, the most relevant diagnostic systems and PFMs (plasma-facing materials) will be described, all accompanied by the corresponding justification for the selection of the materials as well as their main characteristics. Finally, an outlook will be provided on future material development activities to be carried out during the next phase of the conceptual design for DEMO, which is highly dependent on the success of the IFMIF-DONES facility, whose design, operation and objectives are also described in this paper.

/pdf/materials-to-be-used-in-future-magnetic-confinement-fusion-15t6uvbx.pdf

Materials to Be Used in Future Magnetic Confinement Fusion Reactors: A Review

Deuterium (D) retention and surface modifications of hot-rolled polycrystalline tungsten (W) exposed to a low-energy (~40 eV D−1), high-flux (2–5 × 1023 D m−2 s−1) D plasma at temperatures of ~380 K and ~1140 K to fluences up to 1.2 × 1028 D m−2 have been examined by using nuclear reaction analysis, thermal desorption spectroscopy, and scanning electron microscopy. The samples exposed at ~380 K exhibited various types of surface modifications: dome-shaped blister-like structures, stepped flat-topped protrusions, and various types of nanostructures. It was observed that a large fraction of the surface was covered with blisters and protrusions, but their average size and the number density showed almost no fluence dependence. The D depth distributions and total D inventories also barely changed with increasing fluence at ~380 K. A substantial amount of D was retained in the subsurface region, and thickness correlated with the depth where the cavities of blisters and protrusions were located. It is therefore suggested that defects appearing during creation of blisters and protrusions govern the D trapping in the investigated fluence range. In addition, a large number of small cracks was observed on the exposed surfaces, which can serve as fast D release channels towards the surface, resulting in a reduction of the effective D influx into the W bulk. On the samples exposed at ~1140 K no blisters and protrusions were found. However, wave-like and faceted terrace-like structures were formed instead. The concentrations of trapped D were very low (<10−5 at. fr.) after the exposure at ~1140 K.

Deuterium trapping and surface modification of polycrystalline tungsten exposed to high-flux plasmas at high fluences

Nucleation and growth of tungsten nanotendrils grown under divertor-like conditions

Numerical assessment of the new V-shape small-angle slot divertor on DIII-D

The advancement of fusion reactor engineering is currently inhibited by the lack of knowledge surrounding the stability of plasma facing components (PFCs) in a tokamak environment. During normal operation, events of high heat loading occur periodically where large amounts of energy are imparted onto the PFC surface. Concurrently, irradiation by low-energy helium ions present in the fusion plasma can result in the synthesis of a fibre form nanostructure on the PFC surface, called 'fuzz'. In order to understand how this heterogeneous structure evolves and deforms in response to transient heat loading, a pulsed Nd:YAG millisecond laser is used to simulate these events on a fuzz form molybdenum (Mo) surface. Performance was analysed by three metrics: nanostructure evolution, particle emission, and improvement in optical properties. Experiments performed at the upper end of the expected range for type-I edge-localized modes (ELMs) found that the helium-induced nanostructure completely disappears after 200 pulses of the laser at 1.5 MJ m−2. In situ mass loss measurements found that the amount of particles leaving the surface increases as energy density increases and the rate of emission increases with pulse count. Finally, optical properties assisted in providing a qualitative indication of fuzz density on the Mo surface; after 400 pulses at 1.5 MJ m−2, the optical reflectivity of the damaged surface is ~90% of that of a mirror polished Mo sample. These findings provide different results than previous studies done with tungsten (W), and further help illustrate the complicated nature of how transient events of high heat loading in a tokamak environment might impact the performance and lifetime of PFCs in ITER and future DEMO devices (Ueda et al 2014 Fusion Eng. Des. 89 901–6).

https://iopscience.iop.org/article/10.1088/0029-5515/56/3/036005/pdf

Structural response of transient heat loading on a molybdenum surface exposed to low-energy helium ion irradiation

Structural damage due to high flux particle irradiation can result in significant changes to the thermal strength of the plasma facing component surface (PFC) during off-normal events in a tokamak. Low-energy He + ion irradiation of tungsten (W), which is currently the leading candidate material for future PFCs, can result in the development of a fiber form nanostructure, known as “fuzz”. In the current study, mirror-finished W foils were exposed to 100 eV He + ion irradiation at a fluence of 2.6 × 10 24  ions m −2 and a temperature of 1200 K. Then, samples were exposed to two different types of pulsed heat loading meant to replicate type-I edge-localized mode (ELM) heating at varying energy densities and base temperatures. Millisecond (ms) laser exposure done at 1200 K revealed a reduction in fuzz density with increasing energy density due to the conglomeration and local melting of W fibers. At higher energy densities (∼ 1.5 MJ m −2 ), RT exposures resulted in surface cracking, while 1200 K exposures resulted in surface roughening, demonstrating the role of base temperature on the crack formation in W. Electron beam heating presented similar trends in surface morphology evolution; a higher penetration depth led to reduced melt motion and plasticity. In situ mass loss measurements obtained via a quartz crystal microbalance (QCM) found an exponential increase in particle emission for RT exposures, while the prevalence of melting from 1200 K exposures yielded no observable trend.

https://docs.lib.purdue.edu/cgi/viewcontent.cgi?article=1003&context=nepubs

Structural evolution of tungsten surface exposed to sequential low-energy helium ion irradiation and transient heat loading

Silicon carbide (SiC) represents a promising but largely untested plasma-facing material (PFM) for next-step fusion devices. In this work, an analytic mixed-material erosion model is developed by calculating the physical (via SDTrimSP) and chemical (via empirical scalings) sputtering yield from SiC, Si, and C. The Si content in the near-surface SiC layer is predicted to increase during D plasma bombardment due to more efficient physical and chemical sputtering of C relative to Si. Silicon erosion from SiC thereby occurs primarily from sputtering of the enriched Si layer, rather than directly from the SiC itself. SiC coatings on ATJ graphite, manufactured via chemical vapor deposition, were exposed to repeated H-mode plasma discharges in the DIII-D tokamak to test this model. The qualitative trends from analytic modeling are reproduced by the experimental measurements, obtained via spectroscopic inference using the S/XB method. Quantitatively the model slightly under-predicts measured erosion rates, which is attributed to uncertainties in the ion impact angle distribution, as well as the effect of edge-localized modes. After exposure, minimal changes to the macroscopic or microscopic surface morphology of the SiC coatings were observed. Compositional analysis reveals Si enrichment of about 10%, in line with expectations from the erosion model. Extrapolating to a DEMO-type device, an order-of-magnitude decrease in impurity sourcing, and up to a factor of 2 decrease in impurity radiation, is expected with SiC walls, relative to graphite, if low C plasma impurity content can be achieved. These favorable erosion properties motivate further investigations of SiC as a low-Z, non-metallic PFM.

Evaluation of silicon carbide as a divertor armor material in DIII-D H-mode discharges

Impurity transport modeling of the new tungsten (W)-coated, V-shaped small angle slot (SAS) divertor in the DIII-D tokamak was conducted using the SOLPS-ITER plasma edge code package and the DIVIMP impurity tracking code. The inboard baffle of the current SAS divertor will be shifted closer to the outboard baffle, creating a V-corner at the slot vertex. In addition, the outboard baffle will be coated with 10–15 μm of W for experiments studying high-Z sourcing and leakage in a closed divertor. Modeling of the ‘SAS-VW’ divertor predicts that these changes to the inner baffle will reduce W gross erosion by 40× relative to the existing SAS divertor when the outer strike point (OSP) is at the V-corner and the ion B × ∇B drift is towards the divertor, driven primarily by significant cooling near the slot vertex. Most W erosion in SAS-VW is expected to occur near the slot entrance, which may pose a higher risk to core contamination than W eroded deeper in the slot. Adding a new sheath-based prompt redeposition model outlined in Guterl et al (2021 Nucl. Mater. Energy 27 100948) increases the sensitivity of redeposition estimates to near-target plasma conditions and may provide more accurate predictions of net erosion. Moving the OSP outboard from the slot vertex ∼4 cm onto the W-coated region yields a 40× increase in the gross erosion rate and a 50% decrease in the core leakage fraction. Thus slight variations in strike point location may counteract the potential benefits of the tightly-baffled V slot on minimizing erosion. This impurity transport modeling provides useful guidance for future experiments on the SAS-VW divertor focused on high-Z erosion/redeposition, scrape-off layer transport, and core leakage.

Predicting tungsten erosion and leakage properties for the new V-shaped small angle slot divertor in DIII-D

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