Experimental Setup to Investigate Hydrogen Isotope Retention on Powder Samples as Slag/Dust Proxies for Advanced Fusion Reactors

Gamma-ray irradiation effect on deuterium retention in reduced activation ferritic/martensitic steel and ceramic coatings

ARIES-AT magnet systems

This paper presents a conceptual design of the magnet systems for an advanced tokamak fusion reactor (ARIES-AT). The main emphasis of the paper is the extrapolation of the current state-of-the-art in high temperature superconductors and coil design, and their implementation in an advanced commercial fusion reactor concept. A conceptual design of both the poloidal field coils and toroidal field coils is presented. The current design point is described and supported with a preliminary structural analysis and a discussion of the merits, performance, and economics of high temperature vs. low temperature superconductors in an advanced fusion reactor design.

The elucidation of the trapping and detrapping mechanisms of hydrogen isotopes in SiC is one of the most critical issues for future fusion reactors if SiC is used as the first wall and structure material. In this study, 1 keV deuterium (D2+) ions were implanted into SiC and the chemical states of C and Si were evaluated by X-ray photoelectron spectroscopy (XPS). The deuterium desorption and retention were also analyzed by thermal desorption spectroscopy (TDS). The deuterium desorption behavior for SiC was compared to that for Si and graphite, and it was found that deuterium is preferentially trapped by C and, after the saturation of the C-D bond, it is trapped by Si in SiC. Deuterium desorption was found to consist of two stages, namely deuterium desorptions bound to Si and C. Their trapping mechanisms were influenced by the damaged structures produced by the D2+ ion implantation. Finally, deuterium retention in SiC at temperatures above 700 K was higher than that in graphite, indicating that tritium retention in SiC may be high compared to that in graphite during plasma operation.

We present a systematic study that quantifies deuterium (D) retention and ammonia (ND3) production from 316 L stainless steel (SS316L) following the implantation of D ions in conditions similar to the ones expected in the ITER tokamak, i.e. with kinetic energy below 300 eV. Using Temperature Programmed Desorption (TPD) after deuterium ion implantation at 250 eV/D, we show that deuterium retention increases linearly with the D fluence up to 1021 D+m−2, with a retention probability of 18%. For higher D fluence, deuterium retention increases sub-linearly. Analysis of the TPD spectra evolution with varying storage time in vacuum after D implantation, shows that D retention is influenced by D diffusion into the bulk of SS316L. Subsequent to D ion implantation, we evidence the efficient production of ND3 molecules during TPD, between 400 K and 750 K, from the nitrogen present naturally in SS316L. Up to 21% of the D release during TPD can be found in ND3 molecules, indeed. The fraction of ND3 in the total D release depends both on the D ion fluence and the nitrogen concentration profile in the bulk. At least 7% of the D release is found in the form of ND3 molecules, even at a fluence of 2 × 1021 D+m−2 and for a natural N concentration bulk profile. Both N diffusion and D diffusion into the bulk appear to dictate the kinetics of ND3 production. Our findings of efficient production of ND3 in D-implanted austenitic 316 L stainless steel underline the need for similar studies on reduced-activation ferritic/martensitic (RAFM) steels that contain similar content of nitrogen and will be used in fusion reactor prototypes.

Investigation of gas and plasma driven deuterium permeation through a vanadium membrane for particle exhaust application in a fusion reactor

A superconducting magnet system is also one of the important components in an advanced magnetic confinement fusion reactor. Then it is required to have a higher magnetic field property to confine and maintain the steady-state burning deuterium (D)-tritium (T) fusion plasma in the large interspace during the long term operation. Burning plasma is sure to generate 14 MeV fusion neutrons during deuterium-tritium reaction, and fusion neutrons will be streamed and penetrated to superconducting magnet through large ports with damping neutron energy. Therefore, it is necessary to consider carefully not only superconducting property but also neutron irradiation property in superconducting materials for use in a future fusion reactor, and a “low activation and high field superconducting magnet” will be required to realize the fusion power plant beyond International Thermonuclear Experimental Reactor (ITER). V-based superconducting material has a much shorter decay time of induced radioactivity compared with the Nb-based materials. We thought that the V3Ga compound was one of the most promising materials for the “low activation and higher field superconductors” for an advanced fusion reactor. However, the present critical current density (Jc) property of V3Ga compound wire is insufficient for apply to fusion magnet applications. We investigated a new route PIT process using a high Ga content Cu-Ga compound in order to improve the superconducting property of the V3Ga compound wire.

Mechanical properties and microstructural investigations of TIG welded 40 mm and 60 mm thick SS 316L samples for fusion reactor vacuum vessel applications

Deuterium retention studies of an amorphous silicon probe exposed to JIPP T-IIU plasma by thermal desorption spectroscopy

Authors

Liquid metal (LM) divertor concepts explore an alternative solution to the challenging power/particle exhaust issues in future magnetic fusion reactors. Among them, lithium (Li) is the most promising material. Its use has shown important advantages in terms of improved H-mode plasma confinement and heat handling capabilities. In such scenario, a possible combination of tungsten (W) on the first wall and liquid Li on the divertor could be an acceptable solution, but several issues related to material compatibility remain open. In particular, the co-deposition of Li and hydrogen isotopes on W components could increase the associated tritium retention and represent a safety risk, especially if these co-deposits can uncontrollably grow in remote/plasma shadowed zones of the first wall. In this work, the retention of Li and deuterium (D) on tungsten at different surface temperature (200 °C–400 °C) has been studied by exposing W samples to Li evaporation under several D2 gaseous environments. Deuterium retention in the W–Li films has been quantified by using laser induced desorption-mass spectrometry (LID-QMS). Additional techniques as thermal desorption spectroscopy, secondary ion mass spectrometry, profilemetry and flame atomic emission spectroscopy were implemented to corroborate the retention results and for the qualitative and quantitative characterization of the films. The results showed a negligible (below LID sensibility) D uptake at Tsurface = 225 °C, when the W–Li layer is exposed to simultaneous Li evaporation and D2 gas exposition (0.67 Pa). Pre-lithiated samples were also exposed to higher D2 pressures (133.3 Pa) at different temperatures (200 °C–400 °C). A non-linear drastic reduction in the D retention with increasing temperatures was found on the W–Li films, presenting a D/Li atomic ratio at 400 °C lower than 0.1 at.% on a thin film of ≈100 nm thick. These results bode well (in terms of tritium inventory) for the potential utilization of this material combination in a real reactor scenario.

Deuterium transport and retention properties of representative fusion blanket structural materials

Assessment of hydrogen production potential of APEX fusion blanket via cobalt-chlorine and copper-chlorine cycles

Thermal desorption spectrometry of beryllium plasma facing tiles exposed in the JET tokamak

Fabrication of MgB 2 superconducting wires as low activation superconducting materials for an advanced fusion reactor application