Study Of Deuterium Retention Research Articles

Comparative study of deuterium retention and vacancy content of self-ion irradiated tungsten

Self-ion irradiation of pure tungsten with 2 MeV W ions provides a way of simulating microstructures generated by neutron irradiation in tungsten components of a fusion reactor. Transmission electron microscopy (TEM) has been used to characterize defects formed in tungsten samples by ion irradiation. It was found that tungsten irradiated to 0.85 dpa at relatively low temperatures develops a characteristic microstructure dominated by dislocation loops and black dots. The density and size distribution of these defects were estimated. Some of the samples exposed to self-ion irradiation were then implanted with deuterium. Thermal Desorption Spectrometry (TDS) analysis was performed to estimate the deuterium inventory as a function of irradiation damage and deuterium release as a function of temperature. Increase of inventory with increasing irradiation dose followed by slight decrease above 0.1 dpa was found. Application of Positron Annihilation Spectroscopy (PAS) to self-irradiated but not deuterium implanted samples enabled an assessment of the density of irradiation defects as a function of exposure to high-energy ions. The PAS results show that the density of defects saturates at doses in the interval from 0.085 to 0.425 displacements per atom (dpa). These results are discussed in the context of recent theoretical simulations exhibiting the saturation of defect microstructure in the high irradiation exposure limit. The saturation of damage found in PAS agrees with the simulation data described in the paper.

Comparative study of deuterium retention in irradiated Eurofer and Fe–Cr from a new ion implantation materials facility

A new facility to study the interaction of hydrogen isotopes with nuclear fusion-relevant first wall materials, and their retention and release, has been produced. The new facility allows for implanting a range of gases into samples, including tritium. An accurate study of isotope effects, such as the isotopic exchange in damaged microstructure, has previously been difficult due to a background signal of light hydrogen. This new capability will allow virtually background free measurements using tritium and deuterium. The design and build of this facility are described and commissioning results are presented. Within the UKAEA-led tritium retention in controlled and evolving microstructure (TRiCEM) project, this facility is used for the comparative study of deuterium retention in self-ion irradiated Eurofer steel and Fe–Cr alloy. Self-ion bombardment with energies of 0.5 MeV is used to mimic the defects created by neutrons in fusion power plants and the created traps are then filled with deuterium in the new facility. Implanted samples are analysed using thermal desorption spectrometry (TDS), secondary ion mass spectrometry (SIMS), and transmission electron microscopy. Results on the total deuterium content as a function of time, TDS spectra and SIMS analysis are presented. A comparison of the results for Eurofer and Fe–Cr revealed several differences. While some of them may be due to experimental details like different time delays between exposure and analysis, others, such as deuterium retention as function of dose, might be genuine and require further studies.

Study of deuterium retention in Be-W coatings with distinct roughness profiles

Three batches of pure beryllium (Be), pure tungsten (W) and mixed Be:W (50:50) coatings were deposited by using the thermionic vacuum arc (TVA) technique. Different superficial roughness and morphologies were obtained by adjusting the bias voltages in the substrate holder. The coatings were irradiated afterwards with 15keV 2H+ ions up to a fluence of 5e17ion/cm2 with the purpose to investigate the influence of the roughness parameters on the retention of deuterium in Be-W deposits in the absence of other co-deposited impurities. Current densities lower than 0.2W/cm2 were used to avoid the release of deuterium induced by heating loads. The results of the experiment point to the negligible effect of rugosity on the retention behaviour under energetic ion bombardment.

Deuterium retention in tin (Sn) and lithium–tin (Li–Sn) samples exposed to ISTTOK plasmas

The use of lithium (Li) or tin (Sn) as a liquid metal plasma facing component is proposed as a solution to the high power load issue on the divertor region of nuclear fusion reactors. The possibility to use these materials depends on their compatibility with hydrogen plasmas. With the purpose of realizing deuterium retention studies, specimens of pure Sn (99.999% Sn) and Li–Sn alloy (30 at.% Li) were exposed in the ISTTOK edge plasma. Ex situ analysis of the samples was performed by means of ion beam diagnostics. Nuclear reaction analysis (NRA) technique was applied using the D(3He,p)4He reaction to quantify the fuel retention on the samples.In this work the deuterium retention is compared between pure Sn and Li–Sn alloy samples in both liquid and solid states. All the samples were found to have retention ratios smaller than 0.1 at.%. This low retention ratio is expected for pure tin given its high mass and the instability of tin hydrides. However the retention was unexpectedly low for the case of Li–Sn which was thought to be dominated by the lithium fraction in the alloy. These results suggest that tin has a role in the retention mechanism in this material.

The effect of ion flux on plasma-induced modification and deuterium retention in tungsten and tungsten–tantalum alloys

The paper presents the results of an experimental study of deuterium retention in W and W–Ta alloy that were exposed to first-wall relevant low flux (∼1020m−2s−1) deuterium plasma in the ECR plasma generator PlaQ. Subsequent analysis included surface imaging by optical microscopy, deuterium depth profiling by nuclear reaction analysis (NRA) and measurements of deuterium content by thermal desorption spectroscopy (TDS). It was found that under investigated exposure conditions the deuterium content was higher in W–Ta alloy than in W. Combined with the previously reported results showing that under high-flux (∼1024m−2s−1) retention is higher in W instead, this gives rise to a peculiar flux effect – dependence of relative retention between different materials on exposure flux. We interpret this effect as evidence that at different flux ranges different populations of trapping sites determine the retention, namely pre-existing microstructural traps at low-flux exposure and plasma-induced ones at high-flux exposure.

Deuterium retention and surface modifications of nanocrystalline tungsten films exposed to high-flux plasma

Deuterium retention studies are presented for nanostructured tungsten films exposed to high-flux deuterium plasmas. Thin tungsten films of ∼1μm thickness were deposited with pulsed laser deposition (PLD) on bulk tungsten. Surface modifications were studied with scanning electron microscopy and deuterium retention with thermal desorption spectroscopy. Three types of PLD films with different densities and crystallinity were studied after exposure to deuterium plasmas. The surface temperature ranged from about 460K at the periphery to about 520K in the centre of the targets. The films withstand the intense plasma exposure well and maintain their overall integrity. An increase of deuterium retention is observed with decreasing tungsten density and crystallite size. We found that the filling of these thin films with deuterium is significantly faster than for pre-damaged polycrystalline tungsten. We observed formation of micrometer-sized blisters as well as structures on the nanometer scale, both depending on the layer type.

Study of deuterium retention in/release from ITER-relevant Be-containing mixed material layers implanted at elevated temperatures

Abstract D implantation into Be-containing mixed material layers: Be, Be–W (W: ∼6 at.%) and Be–C (C: ∼50 at.%), was performed at elevated temperatures. The temperature dependence of D retention varied depending on the admixed element. D retention in Be and Be–W layers decreases with increasing implantation temperature, while the Be–C layers maintained rather high D retention in the present investigated temperature range (up to 623 K). D desorption behaviour from Be–C suggests the contribution of C–D bonds to D retention. W admixture into Be can significantly suppress D retention in Be. Long-term isothermal annealing at 513 and 623 K for D removal was also performed to simulate the ITER-wall-baking scenario. Even extended annealing at temperatures comparable to or lower than the implantation temperature does not lead to a significant release of retained D.

Design of a planar probe diagnostic system for plasmatron VISIONI and its application for the study of deuterium retention in W–Ta alloys

One of the issues still to be clarified with respect to the possible use of tungsten and tungsten alloys as first wall material in fusion reactors is the retention of hydrogen isotopes under fusion-relevant conditions, such as the presence of helium ash, neutron irradiation, thermal shock. The low flux, high temperature plasma simulator VISIONI, being currently constructed at the Belgian nuclear research center SCK·CEN, will have the unique ability to handle radioactive materials – neutron irradiated samples and tritium plasma. Currently it is being tested with non-irradiated samples and deuterium plasma. To monitor the plasma parameters near the investigated specimen, a planar probe system has been developed. It was demonstrated that the plasmatron is able to provide conditions relevant for the ITER first wall. Investigations of deuterium retention in W–Ta alloys have been performed, using thermodesorption spectroscopy. A hypothesis of the existence of three kinds of trapping sites for deuterium in W–Ta alloys, based on deuterium release studies, is suggested.

Development of key ITER relevant technologies on the Tore Supra tokamak

As the largest superconducting tokamak presently in operation, with all plasma facing components (PFC) actively cooled, Tore Supra is able to routinely address critical issues of high power long duration discharges. Twenty years of operation of Tore Supra have been valorised through “lessons learnt” studies on systems critical for long pulse operation. Real time feedback systems have been developed to safely operate at a high level (>10 MW) of injected power. The versatility of the integrated plasma controller has allowed extending feedbacks to plasma performance optimisation. A dedicated long discharges campaign, cumulating 5 h of plasma in 10 operation days without any conditioning procedure meanwhile, has allowed to gain new insights in deuterium retention studies and to reveal a new operational limit characterised by a progressive decreases of the power handling capability. A strong development effort has been undertaken to follow up the PFC ageing during and between experimental campaigns. This includes the development of an articulated inspection arm, of lock-in thermography and confocal microscopy techniques. An important part of the Tore Supra programme, is the development of reliable heating and current drive systems. Progress in the development of lower hybrid couplers, klystrons and test of a load resilient ion cyclotron range of frequency are reported.

Bulk retention of deuterium in graphites exposed to deuterium plasma at high temperature

A highly ionized deuterium plasma with a low residual gas pressure and a high intensity D 2 +-ion beam were used for the study of deuterium retention in RG-Ti-91 and POCO AXF-5Q graphites. Deuterium retention in the samples was estimated by TDS during heating to 2000 K. Mechanical removal of a surface layer 100 μm thick was used to distinguish bulk and surface fractions of retained deuterium. The samples of RG-Ti and POCO graphites were exposed to a plasma with an ion flux of 3 × 10 17 D/cm 2 · s in the ‘Lenta’ plasma device for 10 to 10 4 s at residual deuterium pressure of 0.04 Pa at 1400 K. Under plasma exposure deuterium capture in RG-Ti graphite reached the saturation level at a fluence of 4 × 10 20 D/cm 2 while the bulk inventory was negligible. As for POCO graphite, deuterium retention increased with fluence and was equal to 18 appm in the bulk for a fluence of 7 × 10 21 D/cm 2. The same amount of deuterium in the bulk was obtained after gas exposure of POCO at an effective pressure of 0.8 Pa (1400 K, 6 h). With this result, the tritium concentration in the plasma-facing graphite materials can reach 1500 appm or 380 grams of tritium per ton of graphite. To understand the role of ion flux in generation of effective pressure, POCO was irradiated with 16 keV D 2 +-ions at 1400 K for 4 h to 8 × 10 20 D/cm 2 (ion flux was 6 × 10 16 D/cm 2 · s, residual deuterium pressure was 0.004 Pa). The results are discussed on the basis of structural differences for POCO and RG-Ti graphites.

ＪＩＰＰＴ‐ＩＩ Ｕプラズマ照射試料中重水素の昇温脱離法による分析

ＪＩＰＰＴ‐ＩＩ Ｕプラズマ照射試料中重水素の昇温脱離法による分析

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

High performance thermal desorption spectroscopy (TDS) has been applied to probe studies of the plasma in JIPP T-IIU. Both analyses of the quantity and the peak temperature of deuterium desorption were made by TDS as a function of radial distance, electron density and ion temperature. Nuclear reaction analysis (NRA) was also provided for radial-profiling of deuterium in the probes. Three amorphous silicon probes were exposed to each discharge plasma in JIPP T-IIU and measured by TDS in a temperature range of room temperature to 973 K and at increments of 10 K/min. The radial distribution of deuterium retention in the amorphous silicon probe obtained by both methods, TDS and NRA, agreed well. No shift of peak temperature was observed with changes of the radial distance, the surface composition of impurities and the quantity of deuterium retained, however, the peak was shifted with the change of ion temperature in the plasma. The result is explained by a simple model from which a relation between the peak temperature of TDS and the projected range of deuterium in the silicon probe is derived. The possibility of energy estimation of the deuterium incident on the silicon probe by the TDS method is proposed.

Biosynthesis of the polyether antibiotic ICI139603 in Streptomyces longisporoflavus: investigation of deuterium retention after incorporation of CD3 13CO2H, 13CD3CO2H, and CH3CD2 13CO2H using 2H n.m.r. and edited 13C n.m.r. spectra

The mechanism of carbocyclic ring formation in the biosynthesis of the polyether antibiotic ICI139603 (2) was investigated by deuterium retention studies after incorporation of CD313CO2H, 13CD3CO2H, and CH3CD213CO2H using 2H n.m.r. spectroscopy, α- and β-isotopic shifts, and edited 13C n.m.r. spectroscopy.