High-heat-flux technologies for the European demo divertor targets: State-of-the-art and a review of the latest testing campaign

You, J.H.; Visca, E.; Barrett, T.; Böswirth, B.; Crescenzi, F.; Domptail, F.; Dose, G.; Fursdon, M.; Gallay, F.; Greuner, H.; Hunger, K.; Lukenskas, A.; Müller, A. von; Richou, M.; Roccella, S.; Vorpahl, C.; Zhang, K.

doi:10.1016/j.jnucmat.2020.152670

Cited by 50 publications

(15 citation statements)

References 40 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…For water-cooled and helium-cooled target designs, W armor is only used as functional material without any structural function and it is acceptable to a certain extent that it operates with a wide temperature range from far below the DBTT to far above RCT. The high heat flux tests have also supported this point [28], which found the fact that the W armor survived 20 MW m −2 test cycles without forming any discernible cracks even though the upper half of the armor has been fully recrystallized and the surface layer has undergone abnormal grain growth. Once recrystallization is accepted, a much wider operational window will be available for the designers.…”

Section: Structure Designmentioning

confidence: 70%

A novel liquid lithium jet-cooled finger-type divertor target concept for fusion power plant application

et al. 2021

View full text Add to dashboard Cite

The divertor target is the most thermally loaded plasma-facing component in a foreseen DEMO reactor and beyond, which has to tolerate the peak high heat fluxes of up to ∼20 MW m−2 produced by intense plasma bombardment, radiation and nuclear heating. However, none of current designs including water-cooled and helium-cooled concepts can satisfy this requirement. Motivated by the excellent power removal capacity of liquid metal coolant and combined with the structure characteristics of the finger-type helium-cooled target, a novel concept of liquid Li jet-cooled finger-type divertor target for DEMO reactors was proposed in this paper. The performance analysis, including thermal-hydraulics analysis, mechanical analysis and MHD effects analysis, have shown that the proposed design can withstand 20 MW m−2 heat load because the temperatures of the structural materials remain within the thermal rules and the maximum thermo-mechanical stress in the VM-W thimble is approximately 484 MPa appearing in the round corner, which is below the 3S m limit at the corresponding temperature. Moreover, a theoretical and empirical analysis has confirmed that MHD effects on pressure drop and heat transfer is rather limited in the design. The comparison of this new design with other representative designs including water-cooled ITER-like target design and helium-cooled modular jet target design has been made, and the results shows that the proposed liquid Li cooled target design has better performance under 20 MW m−2 high heat flux and ∼10 dpa neutron irradiation. Therefore, this design is promising to provide a new option for solving the DEMO reactor divertor heat removal issues. Certainly, a large number of R&D efforts are still needed to ensure the success of this concept, particularly in the areas of materials, fabrication and irradiation.

show abstract

Section: Structure Designmentioning

confidence: 70%

A novel liquid lithium jet-cooled finger-type divertor target concept for fusion power plant application

et al. 2021

View full text Add to dashboard Cite

show abstract

“…irradiation creep above 350 • C and irradiation embrittlement below 250 • C [32]. Thus, it is reinforced by various kinds of Cu-W composite materials and novel interlayer materials as shown in figure 11 [41].…”

Section: Development Of Target Technologies For Demomentioning

confidence: 99%

Power exhaust concepts and divertor designs for Japanese and European DEMO fusion reactors

et al. 2021

Self Cite

View full text Add to dashboard Cite

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan–Europe broader approach DEMO design activity (BA DDA). Common critical issues are the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling (P rad tot/P heat ⩾ 0.8). Different exhaust concepts in the main plasma and divertor have been developed for Japanese (JA) and European (EU) DEMOs. JA proposed a conventional closed divertor geometry to challenge large P sep/R p handling of 30–35 MW m−1 in order to maintain the radiation fraction in the main plasma at the ITER-level (f rad main = P rad main/P heat ∼ 0.4) and higher plasma performance. EU challenged both increasing f rad main to ∼0.65 and handling the ITER-level P sep/R p in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding P sep = 250–300 MW and 150–200 MW, respectively. Both results showed that large divertor radiation fraction (P rad div/P sep ⩾ 0.8) was required to reduce both peak q target (⩽10 MW m−2) and T e,i div. In addition, the JA divertor performance with EU-reference P sep of 150 MW showed benefit of the closed geometry to reduce the peak q target and T e,i div near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130 °C–200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs.

show abstract

“…For viable commercial D-T nuclear fusion, it is essential that the breeding of tritium at least balances the tritium which is retained after having penetrated the components. One of the critical reactor components requiring study in this regard is the tungsten divertor, which is anticipated to undergo very high neutron flux [1,2] and so develop a highly damaged microstructure, especially near coolant pipes where the temperature is kept below the onset of stage III recovery (vacancy migration) [3][4][5][6]. It is well established that irradiation-induced * Author to whom any correspondence should be addressed.…”

Section: Introductionmentioning

confidence: 99%

An empirical potential for simulating hydrogen isotope retention in highly irradiated tungsten

Mason,

Nguyen-Manh,

Lindblad

et al. 2023

J. Phys.: Condens. Matter

View full text Add to dashboard Cite

We describe the parameterization of a tungsten-hydrogen empirical potential designed for use with large-scale molecular dynamics simulations of highly irradiated tungsten containing hydrogen isotope atoms, and report test results. Particular attention has been paid to getting good elastic properties, including the relaxation volumes of small defect clusters, and to the interaction energy between hydrogen isotopes and typical irradiation-induced defects in tungsten. We conclude that the energy ordering of defects changes with the ratio of H atoms to point defects, indicating that this potential is suitable for exploring mechanisms of trap mutation, including vacancy loop to plate-like void transformations.

show abstract

High-heat-flux technologies for the European demo divertor targets: State-of-the-art and a review of the latest testing campaign

Cited by 50 publications

References 40 publications

A novel liquid lithium jet-cooled finger-type divertor target concept for fusion power plant application

A novel liquid lithium jet-cooled finger-type divertor target concept for fusion power plant application

Power exhaust concepts and divertor designs for Japanese and European DEMO fusion reactors

An empirical potential for simulating hydrogen isotope retention in highly irradiated tungsten

Contact Info

Product

Resources

About