Hydrogen Isotope Retention Research Articles

Effect of impurity deposition layer formation on D retention in LHD plasma exposed W

Effect of carbon based mixed-material deposition layer formation on hydrogen isotope retention was studied. The tungsten (W) samples were placed at four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (higher heat load area), and ER (erosion dominated area) during 2013 plasma experimental campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼ 4000 shots of hydrogen plasma in a 2013 plasma experimental campaign. Most of the sample surface except for ER was covered by a mixed-material deposition layer formed by plasma experimental campaign, which consisted of carbon, but some metal impurities were contained. For ER sample, He bubbles were formed due to long term He discharge cleaning and He plasma experiments during the plasma experimental campaign. The additional 1keV D2+ implantation was performed to evaluated the D retention enhancement by plasma exposure. It was found that both of H and D retentions were clearly increased. In particular, the H retention was controlled by the thickness of the carbon-dominated mixed-material deposition layer, indicating most of the H was trapped by this mixed-material deposition layer. It is concluded that the accumulation of low-Z mixed-material layer on the surface of the first wall is one of key issues for the determination of hydrogen isotope retention in first wall.

Fusion related research with laser-induced-breakdown-spectroscopy on metallic samples at the ENEA-Frascati laboratory.

The study of plasma-wall interactions is of paramount importance for continuous and fault free operations in thermonuclear fusion research to monitor the damages of plasma facing components (PFCs), plasma pollution from impurities and wall retention of hydrogen isotopes, like tritium. These needs make laser-induced-breakdown-spectroscopy (LIBS) a suitable candidate for a real time monitoring of PFCs in the current and next generation fusion devices, like ITER. It is also worthwhile for the quantitative analysis of surfaces, with micro-destructivity of the sample and depth profiling capabilities with sub-micrometric sensitivity. In this paper LIBS spectroscopy is exploited as a valid diagnostic tool for PFCs at the ENEA Research Center in Frascati (Italy) and at the Institute of Plasma Physics and Laser Microfusion (IPPLM) of Warsaw (Poland). The activities have been focused on LIBS characterization of samples simulating PFCs surfaces eroded/redeposited or contaminated from nuclear fuel after or during the normal operation of the reactor.

In situ monitoring hydrogen isotope retention in ITER first wall

Tritium retention inside the vacuum vessel is a potentially serious constraint in the operation of large-scale fusion machines like ITER. An in situ diagnostics for first wall H/D/T retention by laser induced desorption spectroscopy (LIDS) is proposed for use between plasma discharges. The technique is based on local baking of the first wall by laser irradiation and subsequent analysis of the in-vessel gas by optical emission spectroscopy of plasma radiation. The local heating implementation, kinetics of H/D/T thermal extraction and the accuracy of optical emission spectroscopy measurements are analysed. To resolve the H/D/T lines spectroscopically, their thermal broadening should be minimized to prevent overlapping of the line shapes. A comparative performance analysis of several types of plasma sources with relatively cold ions is made including the following types of discharges: Penning, RF multipactor, laser torch and ECR. All these radiation sources require rather low power and could be used for remote in situ measurements of relative densities of the thermally extracted hydrogen isotopes.

Determination of 9Be(p,p0)9Be, 9Be(p,d0)8Be and 9Be(p,α0)6Li cross sections at 150° in the energy range 0.5–2.35 MeV

Beryllium and Tungsten have been chosen as the new “plasma facing materials” for fusion reactors. Understanding plasma wall interactions and their effects such as erosion and redeposition leading to the formation of alloys from the different materials present in the reactor chamber, is critical to model the retention of hydrogen isotopes by the wall materials. Determining the amount and the depth profile of Be and other elements deposited in the walls is a mandatory requirement in fusion reactors, making the knowledge of the relevant cross sections essential for IBA analysis. In this work we measured the 9Be(p,p0)9Be, 9Be(p,d0)8Be and 9Be(p,α0)6Li cross sections in the energy range 0.5–2.35MeV, at an angle of 150°.

Quantitative Microstructure and Defect Density Analysis of Polycrystalline Tungsten Reference Samples after Different Heat Treatments

Abstract In order to provide a solid basis for the correlation of microstructure and hydrogen isotope retention in tungsten, reference samples with different microstructures were prepared from a single batch of polycrystalline tungsten by standardised polishing and heat treatment procedures. Representative samples were analysed by scanning electron microscopy and scanning transmission electron microscopy as well as by electron backscatter diffraction. We show that if the annealing temperature is increased from 1 200 to 1 500 K, practically only the density of dislocations and grain boundaries with very small misorientations of less than 2° is reduced, while for annealing at 1 700 and 2 000 K, also the density of high-angle grain boundaries is reduced due to grain growth. Furthermore, the dislocation density is reduced by nearly two orders of magnitude compared to tungsten annealed at 1 200 K. We also comment on two different textures on the front and rear side of the samples that were observed both by X-ray diffraction and EBSD.

Predicting hydrogen isotope inventory in plasma-facing components during normal and abnormal operations in fusion devices

Hydrogen isotope behavior and inventory in plasma-facing components (PFCs) of fusion devices are key concerns for safe, reliable, and economical operation. To accurately estimate hydrogen isotope retention and recovery in tungsten (the current leading candidate as a PFC), we have developed a model that was recently benchmarked against isotope depth profile and retention level in a tungsten target under various conditions and compared with both experimental data and simulation results. In this research, we have extended the model to include details of transient events. Therefore, one can use this model to estimate hydrogen isotope retention behavior in tungsten and potential other PFC candidates during normal operational pulse, effects of edge-localized modes (ELMs), and a possible cleaning processes scenario.

Macroscopic rate equation modeling of trapping/detrapping of hydrogen isotopes in tungsten materials

Relevant parameters for trapping of Hydrogen Isotopes (HIs) in polycrystalline tungsten are determined with the MHIMS code (Migration of Hydrogen Isotopes in MaterialS) which is used to reproduce Thermal Desorption Spectrometry experiments. Three types of traps are found: two intrinsic traps (detrapping energy of 0.87 eV and 1.00 eV) and one extrinsic trap created by ion irradiation (detrapping energy of 1.50 eV). Then MHIMS is used to simulate HIs retention at different fluences and different implantation temperatures. Simulation results agree well with experimental data. It is shown that at 300 K the retention is limited by diffusion in the bulk. For implantation temperatures above 500 K, the retention is limited by trap creation processes. Above 600 K, the retention drops by two orders of magnitude as compared to the retention at 300 K. With the determined detrapping energies, HIs outgassing at room temperature is predicted. After ions implantation at 300 K, 45% of the initial retention is lost to vacuum in 300 000 s while during this time the remaining trapped HIs diffuse twice as deep into the bulk.

Preheating Temperature Effect on Tritium Retention in VPS-W

The deuterium retention behavior for the Vacuum Plasma Spraying (VPS) tungsten (W) coating was studied to demonstrate the tritium retention as a function of heating temperature. It was found that two major deuterium desorption stages were observed at the temperature regions of 400 - 700 K (Stage 1) and 900 - 1100 K (Stage 2), considering that Stage 1 was linked to the desorption of deuterium trapped by near surface and intrinsic defects, and Stage 2 was related to the desorption of deuterium bound to impurities as C-D bonds. By heating the sample above 673 K, the major peak of C-1s was shifted from C-O bond to C-C bond, where the retention of deuterium as Stage 2 was increased. Therefore it was indicating that the hydrogen isotope retention was controlled by the amount of C-C bond in VPS, most of which was contaminated during the VPS coating process. The comparison of several samples (VPS-W with shading, VPS-W without shading and Polycrystalline W (PCW)) shows that the carbon impurity has a large affinity with deuterium and make stable trapping states compared to that with intrinsic defects and grain boundaries. However, most of them was reduced by heating at 1173 K. Therefore, heating treatment is quite important to get rid of carbon impurities and refrain higher tritium retention in VPS.

Suppression of hydrogenated carbon film deposition and hydrogen isotope retention by nitrogen addition into cold remote H/D and CH4 mixture plasmas

Control of tritium retention and its removal from the first wall of future fusion devices are one of the most crucial issues for safety and effective use for fuel. Nitrogen addition into remote edge plasmas has been considered and tested as an effective method for suppression of carbon film deposition and reduction of hydrogen isotope absorption in the deposited films. In this paper we have investigated the scavenger effects of nitrogen injected into low temperature D2/CH4 plasmas on hydrogenated carbon film growth using a small helical device. The result of the deposition shows that the key reactive particles with CN and ND(H) bonds to suppression of hydrogenated carbon film growth and hydrogen isotope absorption are much slowly generated compared with hydrocarbon particles such as CD(H)x and C2D(H)x. This may be due to the slow atomic nitrogen diffusion into hydrogenated carbon layer and the chemical equilibrium between nitrogen absorption.

Thermodynamics of hydrogen-induced superabundant vacancy in tungsten

We investigate superabundant vacancy formation induced by hydrogen in tungsten in terms of an equilibrium thermodynamic model to estimate hydrogen isotope retention in plasma facing materials. Vacancy-hydrogen cluster concentrations in the bulk tungsten are calculated as a function of the H concentration at finite temperature. A monovacancy in usual bcc transition metals is capable of accommodating six H atoms, while a maximum of 12 H atoms can be accommodated in a tungsten monovacancy, according to first-principle calculations. The present results provide thermodynamic profiles of vacancy-hydrogen clusters trapping more than six H atoms for the first time. In present work, configurational transitions of H atoms trapped in the monovacancy and activation energies for them are investigated by examining the transition paths in order to calculate configurational entropy. Vacancy-hydrogen clusters trapping more than six H atoms exist in thermodynamic equilibrium. However, the major vacancy-hydrogen clusters are composed of six H atoms in a wide range of temperature and H concentration.

Retention of deuterium in damaged low-activation steel Rusfer (EK-181) after gas and plasma exposure

Reduced-activation ferritic–martensitic steels (RAFMS) are advanced structural materials for the construction of future fusion reactors with high fluxes of neutrons, such as DEMO or a Fusion Neutron Source (FNS).In the present work the influence of different damages on deuterium retention in the RAFM Rusfer (Chernov et al., 2007) was investigated. Three different types of damage were applied:(i)irradiation by 20MeV W6+ ions to a damage fluence of 0.89dpa (1.4×1018ions/m2). Tungsten ion irradiation was used as proxy for displacement damage created by neutrons;(ii)heat loads in the QSPA-T facility with 10 pulses of 0.5MJ/m2 with a duration of 0.5ms;(iii)low-temperature hydrogen plasma irradiation in the LENTA facility at 320 and 600K to a fluence of 1025H/m2.The hydrogen isotope retention properties of the damaged and undamaged material were investigated by exposure to deuterium gas several weeks after damaging. The deuterium retention was investigated in the temperature range of RT–773K at pressure 104Pa. Deuterium depth profiles were measured a month after gas exposure by nuclear reaction analysis (NRA) using the D(3He,p)α nuclear reaction. Deuterium retention in damaged and undamaged Rusfer in the temperature range of 300–600K has a maximum at 500K for all types of damage investigated. The typical value of deuterium concentration in the bulk is 10−3at.%. Peculiarities of D retention in damaged samples are discussed.

Effect of induced damage on hydrogen isotope retention of F82H with impurity layer

The effects of induced damage on hydrogen isotope retention in F82H with or without thermal oxidation were investigated using thermal desorption spectroscopy. To induce damage and modify the surface, glow discharge pre-irradiated Ar+ ions was examined. In non-oxidized samples, the amount of desorbed deuterium increased with Ar+ ion fluence. Oxygen depletion in the surface layer of non-oxidized samples from the Ar+ ion irradiation, which resulted in bulk diffusion of deuterium, is responsible for the increase in deuterium retention. A comparison between non-oxidized and oxidized samples clearly indicated that the surface oxide layer greatly influenced deuterium retention/desorption behaviors of F82H.

Multiscale modelling of plasma–wall interactions in fusion reactor conditions

The interaction of fusion reactor plasma with the material of the first wall involves a complex multitude of interlinked physical and chemical effects. Hence, modern theoretical treatment of it relies to a large extent on multiscale modelling, i.e. using different kinds of simulation approaches suitable for different length and time scales in connection with each other. In this review article, we overview briefly the physics and chemistry of plasma–wall interactions in tokamak-like fusion reactors, and present some of the most commonly used material simulation approaches relevant for the topic. We also give summaries of recent multiscale modelling studies of the effects of fusion plasma on the modification of the materials of the first wall, especially on swift chemical sputtering, mixed material formation and hydrogen isotope retention in tungsten.

A First Simulation Study of Hydrogen Isotope Accumulation and Diffusion in Tungsten First Wall of HiPER Inertial Fusion Facility

The study of the retention and desorption of hydrogen isotopes and helium atoms in first-wall materials is key for the design of future fusion reactors, not only for the effect of the materials on the degradation of the wall properties but also for the implications in tritium management strategies.A diffusion model of the implanted H, D, T, and He species in a 1-mm-thick first wall of tungsten for the two initial phases of the proposed European laser fusion project HiPER (namely, phases 4a and 4b) has been implemented using the tritium migration analysis program TMAP7. The effects of the abrupt temperature increases, working temperatures, and the operational pulsing modes on the diffusion are studied. Although a detailed treatment of the different trapping mechanisms has been omitted, meaningful quantitative results on the accumulation, desorption, and time intervals to reach a stationary state are presented and discussed.

Development of a plasma driven permeation experiment for TPE

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800°C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000°C, a new TPE membrane holder has been built to hold test specimens (≤1mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000°C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700Pa) with good flow uniformity (at 1000sccm). Thermal transport simulations indicate that temperatures up to 1000°C are expected at the highest TPE fluxes.

Three-dimensional microscopy of deuterium in tungsten

The hydrogen isotope retention in tungsten is an important issue for fusion devices. In this paper we study the possibility of using a μm-focused deuterium beam in order to quantify deuterium distributions in microscopic dimensions. Due to the lack of cross-section data for deuteron–deuteron-scattering (dd-scattering) a validated reference sample is needed. For this purpose we used a 15 μm thick aluminum foil covered by a-C:D-layers that have been deposited in a CD4 plasma discharge from both sides. At the SNAKE facility of the Munich tandem accelerator we already established a three-dimensional microscopy of hydrogen using protons within an energy range between 17 and 25 MeV. Now, we have tested the possibility for deuteron microscopy. As a first application we analyzed a 25 μm foil implanted with 2.0 × 1020 atom cm−2 deuterons.

Current status of nanostructured tungsten-based materials development

Nanostructured tungsten (W)-based materials offer many advantages for use as plasma facing materials and components exposed to heavy thermal loads combined with irradiation with high-energy neutron and low-energy ion. This paper first presents the recent progress in nanostructured toughened, fine grained, recrystallized W materials. Thermal desorption spectrometry apparatus equipped with an ion gun has been installed in the radiation controlled area in our Center at Tohoku University to systematically investigate the effects of displacement damage due to high-energy neutron irradiation on hydrogen isotope retention in connection with the nano- or micro-structures in W-based materials. In this paper, the effects of high-energy heavy ion irradiation on deuterium retention in W with different microstructures are described as a preliminary work with the prospective view of neutron irradiation effects.

Dislocations mediate hydrogen retention in tungsten

In this letter, a comprehensive mechanism for the nucleation and growth of bubbles on dislocations under plasma exposure of tungsten is proposed. The mechanism reconciles long-standing experimental observations of hydrogen isotopes retention, essentially defined by material microstructure, and so far not fully explained. Hence, this work provides an important link to unify material's modelling with experimental assessment of W and W-based alloys as candidates for plasma facing components.

Modelling of hydrogen isotope retention in the tungsten divertor of EAST during ELMy H-mode

In this work, we study hydrogen isotopes (HI) inventory inside tungsten plasma-facing materials during high confinement mode discharges with repetitive edge localized modes (ELMy H-mode) based on the operating parameters of the EAST device, since tungsten is considered as the primary plasma-facing material and the ELMy H-mode is an important operation regime for EAST and future devices. The upgraded Hydrogen Isotope Inventory Processes Code (HIIPC) is applied with the incident depth profile provided by SRIM-2013 to make the study. The code is first verified by comparison with experimental measurements. The effects of the incident ion energy and ion flux on the retention are then studied. Finally, using the parameters obtained from EAST diagnostics, the HI retention inside the W divertor during ELMy H-mode is studied, which indicates the retained HI can be increased dramatically mainly due to ion-induced trap sites by ELMs.

Comparison of hydrogen isotope retention for tungsten probes in LHD vacuum vessel during the experimental campaigns in 2011 and 2012

To evaluate hydrogen isotope retention enhancement in W by plasma exposure, the stress relieved tungsten samples were placed at three or four different positions, namely PI (sputtering erosion dominated area), DP (deposition dominated area), HL (Higher heat load area) and ER (erosion dominated area) during 2011 (15th) or 2012 (16th) plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ∼6700 shots of hydrogen plasma in a 2011 plasma experiment campaign and ∼5000 shots in a 2012 plasma campaign. Thereafter, additional 1.0keV deuterium ion implantation was performed to evaluate the change of hydrogen isotope retention capacity by plasma exposure. It was found that more than 50 times of hydrogen retention enhancement for DP sample was derived compared to that for pure W. In especially, the carbon-dominated mixed-material layer would control the hydrogen isotope retention for all the area except for erosion-dominated area, indicating that the chemical structure for carbon-dominant mixed-material layer would govern the H and D retention enhancement for most area by long-term plasma exposure. Therefore, the surface area for carbon material would be one of key issues for the determination of hydrogen isotope retention in first wall, even if all tungsten first walls will be used.