Comparison of passivation behavior of SS316L with that of SS304 in tritiated water solution

The effects of tritium on the passivation behavior and passive layer formed in tritiated water circum- stance for SS316L were investigated by means of an anodic polarization measurement technique and X- ray photoelectron spectroscopy, respectively. The results were compared with those for SS304, since it was predicted from a model of the tritium effects on corrosion suggested in the previous studies that SS316 would be less affective to tritiated water circumstance than SS304. As the results, the passiva- tion inhibitory effect of tritium could not be observed for SS316L, while it was observed for SS304 and the other researched materials so far, as predicted. However, the thickness of the passive layer and the boundary between the passive layer and bulk of SS316L were found affected by tritium; thickened and gradated, respectively. From these results, it was concluded that SS316L would be more sustainable in tri- tiated water circumstance than SS304, although the corrosion of SS316L would be more or less enhanced in tritiated water circumstance

Assessment on liquid Li fire risk under humid air condition and with heat insulators

The IFMIF-like fusion neutron sources (FNSs) such as A-FNS and IFMIF-DONES have Li fire risks and the R&D activity for the Li fire risk reduction has been conducted under the IFMIF/EVEDA project. The objectives of this study are to experimentally elucidate humidity condition in air at which no Li fire ignites and the effect of the heat insulator on the Li fire ignition. The experimental set-up was designed and fabricated in order to perform Li fire experiments safely and systematically. Around 1 g of metallic Li samples were heated with and without the heat insulators as an experimental parameter of humidity in air. In this work, no ignition was observed under the less than 0.15 vol% humidity air. The heating experiments with the heat insulators clarified that the reaction between Li and the heat insulators causing Li fire ignition didn’t occur. In conclusion, the results of the present study suggest the conditions of less than 0.15 vol% humidity air and employing the heat insulator of alkaline earth silicate wool are useful for the Li target system of FNSs

Impact of the beam pressure on the free surface of the liquid lithium target of fusion neutron sources

Accelerator-driven fusion neutron sources, e.g. Advanced Fusion Neutron Source (A-FNS) and DEMO Oriented Neutron Source (IFMIF-DONES), are being designed, based on the achievements and lessons learnt from the IFMIF/EVEDA project. The purpose of the study is to analytically evaluate the deformation of the lithium free surface, potentially created by the intense beam injection. We have developed a two-dimensional (2-D) fluid model of the “beam-on” target of an IFMIF-type fusion neutron source by extending the phenomenological 2-D model of the Li target flow of the EVEDA Lithium Test Loop (ELTL). We have derived, from the newly developed model, a practical and analytical formula of the thickness of the “beam-on” target in the flow direction with accounting for the beam injection effect. We have also evaluated the thickness of the beam-on target of A-FNS as an example of application of the derived formula. The evaluation indicates that under the reference design conditions of the accelerator and beam-on Li target of A-FNS, the deformation of the free surface of the target by the beam injection is negligibly small. The analytical results presented here may be used for a benchmark of more complicated CFD simulation for the “beam-on” target design of IFMIF-type fusion neutron sources like A-FNS and IFMIF-DONES. Keywords: IFMIF-type fusion neutron source, Design, Liquid lithium target, Beam injection, Free surface, Modelin

Deuterium permeation behavior for damaged tungsten by ion implantation

The deuterium (D) permeation behaviors for ion-damaged tungsten (W) by 3 keV D-2 (+) and 10 keV C+ were studied. The D permeability was obtained for un-damaged W at various temperatures. For both D-2 (+) and C+ implanted W, the permeability was clearly reduced. But, for the D-2 (+) implanted W, the permeability was recovered by heating at 1173 K and it was almost consistent with that for un-damaged W. In the case of C+ implanted , the permeability was not recovered even if the sample was heated at 1173 K, indicating that the existence of carbon would prevent the recovery of permeation path in W. In addition, transmission electron microscope (TEM) observation showed the voids were grown by heating at 1173 K and not removed, showing the existence of damages would not largely influence on the hydrogen permeation behavior in W in the present study

Correlation of surface chemical states with hydrogen isotope retention in divertor tiles of JET with ITER-Like Wall

To understand the fuel retention mechanism correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011–2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat loadPeer reviewe

Retention behavior of hydrogen isotopes in boron film deposited on SS-316 for LHD first wall

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Comparison of Hydrogen Isotope Retention in Divertor Tiles of JET with the ITER-Like Wall Following Campaigns in 2011–2012 and 2015–2016

Hydrogen isotope retention and chemical state for the tiles exposed to plasma in the JET–ITER-like wall (ILW) during two campaigns in 2011–2012 (first campaign, ILW-1) and 2015–2016 (third campaign, ILW-3) were studied and compared by means of X-ray photoelectron spectroscopy and thermal desorption spectroscopy. In both campaigns the upper part of the inner divertor tiles was the deposition-dominated area, while erosion was observed on the outer divertor tiles. Therefore, higher deuterium retention was found on the inner divertor tiles. The major D desorption peak for the inner divertor tiles from ILW-3 was located at the temperature range of 470°C to 520°C, which was higher than measured after ILW-1: 370°C to 430°C. The XPS analyses showed the formation of a BeO layer on the ILW-3 inner divertor tiles, while after ILW-1 the layers also contained a significant amount of carbon. Deuterium retention was reduced toward the outer divertor tiles. The differences could be related to the difference in the power level in the two campaigns