Deuterium Inventory Research Articles

Deuterium retention in helium-irradiated tungsten

Hot-rolled tungsten samples were first irradiated with helium ions with varying fluences or energies and then exposed to deuterium plasmas, and the influence of helium fluence (corresponding to damage levels of 0.12–2.3 dpa) and helium ion incident energy (50, 200, 400 keV) on deuterium retention were investigated. A significant increase of the total deuterium inventory was observed in the helium irradiated sample, and deuterium retention reaches saturation at 0.69–1.15 dpa. Radiation-induced defects and helium-vacancy complexes or helium bubbles contribute to the increase of deuterium inventory, and the effect of helium-vacancy complexes or helium bubbles is aggrandizement with the increase of damage level. Comparison of deuterium depth profiles of samples irradiated by helium ions with different incident energies reveals that the vast majority of deuterium is trapped in the helium accumulation region, and the depth of the maximum deuterium concentrations is consistent with the maximum helium concentrations rather than the damage levels. Furthermore, the total amount of deuterium increases with the increase of helium ions energy due to the deepening of the irradiation region.

Vacuum Pumping System for ITER's Deuterium Tritium Operational Campaign

SummaryIn the ITER fusion experiment the torus vacuum pumping system is mainly involved in pumping plasma exhaust gases consisting of Deuterium/Tritium (DT) and trace quantity of helium. The currently proposed plasma operation plan for the six torus cryopumps is staggered for all fueling rate and plasma durations and therefore variable tritium inventories need to be handled. The ITER cryopumps and the backing vacuum pumping system are custom made tritium compatible pumping trains. Their compliant matrix not only asks for pumping efficiencies and hydrogen confinement but also respecting the specific tritium and deuterium (DT) inventories for measures on radiation protection and fire safety. In this article, we present a possible operation plan for torus vacuum pumping system in combinations with roughing pump trains for the ITER baseline plasma scenario. The operation of the vacuum system is driven by DT inventories in the different pumps and building sectors.The primary vacuum pump is torus cryogenic pump (TCP) followed backing roughing pump trains comprise the cryogenic viscous flow compressor (CVC), tritium compatible scroll and Roots pumps. The TCPs are regenerated during timed intervals for an uninterrupted and safe torus operation. Next to the demanding long pulse operation (1 hour) we present an operation plan for inductive short pulse (450 s), hybrid pulse (1000 s) and dwell time vacuum pumping operations [1]. However, the inductive and hybrid pulses are not as critical as long pulse operation pertaining to tritium and hydrogen inventory management.

Diffusivity of hydrogen and properties of point defects in beryllium investigated by DFT

Beryllium will be one of the plasma-facing materials for ITER. It will have to sustain high fluxes of hydrogen isotopes and as a consequence significant amounts of tritium can be retained in the wall. For safety and operational reasons, the deuterium and tritium inventory in the vacuum vessel must be limited. As a consequence, hydrogen diffusion, trapping and solubility are of vital importance in assessing and modeling the plasma fuel retention into the wall. In order to understand these issues, point defects and the behavior of hydrogen in beryllium are investigated based on Density Functional Theory calculations. Although some data have already been acquired in the past, some of them disagree, which motivates further investigations. To do so, the formation energy and diffusion properties of point defects are investigated in the first part of this paper. In a second part, the solubility and diffusivity of hydrogen in beryllium are calculated. A diffusion coefficient is established in order to be used in Rate-Equation and Kinetic Monte-Carlo Kinetic models and to allow for comparison with experimental measurements.

Radiative detached divertor with acceptable separatrix Zeff

The feasibility study is performed for the radiative detached divertor (RDD) concept, which characterizes a variety of detached plasmas with impurity amounts providing the highest levels of divertor radiation without X-point MARFE (XPM), based on a set of restrictive criteria for the leading plasma parameters (LPPs) including, e.g., tolerable peak power loads below 1 MW/m2, low separatrix densities, allowable impurity concentrations, and acceptable Zeff values for a DIII-D like tokamak. For this, extensive simulations with the 2-D edge plasma transport code were done scanning the deuterium and impurity inventories practically for all impurity elements from beryllium to neon and the analysis of LPP variations in these scans is presented. It is shown that, for a given D inventory, the total radiation fraction with an increase in the impurity inventory reaches a flat top level, frad = 0.85 ± 0.01, whereas the higher frad corresponds to XPM. This critical fraction is the same for all elements and values of the D inventory. Successful RDD solutions with a flat top radiation meeting all ad hoc LPP criteria are found for some elements. Boron and nitrogen are shown to be the most promising elements for seeding, since they are capable of providing alone the successful RDD at the lowest concentrations. Several important effects on impurity radiation are considered including: cross-field impurity transport in regions with strong temperature gradients, multi-species thermal force, charge-exchange of impurity ions with D atoms originating from recombination, impurity entrainment by parallel flows, flows caused by inner/outer divertor asymmetries, and Mach ∼ 1 flows reached inside radiation-ionization fronts. The impurity radiation profiles of various elements are analyzed suggesting three patterns differing in the radiation front position with respect to the D ionization source. The modeled relocation of D from the pedestal into divertor regions; an enhanced pedestal enrichment for impurities with high first ionization potential; and RDD capability of confining inside the divertors the large D inventories as neutral atoms, are considered as effects of increasing impurity inventory. The responses of a flat-top RDD to the input parameters variation are studied. For instance, dependencies of LPPs on the input power increase, which express the sharp and even bifurcation transitions between two distinct branches corresponding to the detached and attached divertors, are discussed.

The high-flux effect on deuterium retention in TiC and TaC doped tungsten at high temperatures

Samples made of tungsten (W) doped either with titanium carbide (W-1.1TiC) or tantalum carbide (W-3.3TaC) were exposed to a low-energy (40 eV/D), high-flux (1.8–5 × 1023 D/m2s) deuterium (D) plasma at temperatures of about800 K, 1050 K, and 1250 K to a fluence of about1 × 1027 D/m2. The deuterium (D) inventory in the samples was examined by nuclear reaction analysis and thermal desorption spectroscopy. At 800 K the D bulk concentrations and total D inventories in W-1.1TiC and W-3.3TaC were more than one order of magnitude higher compared to that in pure polycrystalline W. At 1050 K and 1250 K the D concentrations in all types of samples were very low (≤10−5 at. fr.); however the D inventories in W-1.1TiC were significantly higher compared to those in W-3.3TaC and pure W. It is suggested that D trapping inside the carbide precipitates and at their boundaries is essential at high temperatures and high incident fluxes, especially in W-1.1TiC.

Deuterium retention in RAFM steels after high fluence plasma exposure

Deuterium retention and detrapping behavior in the ferritic-martensitic steels EUROFER’97 and P92 after exposure to plasma at high fluences ≥ 1026D/m2 was studied using thermal desorption spectroscopy (TDS), supported by nuclear reaction analysis. Low-temperature irradiation at 450K and fluences ≥ 1026D/m2 with low impact energy D+ / D++He+ ions of 40eV at PSI-2 resulted in a deuterium inventory of 7–18 ×1019D/m2 predominantly at depths ≥8.6µm. Helium admixture led to a reduction of total D retention in both steels, irrespective of surface erosion and composition. The deuterium spectra of both steels displayed one D2 desorption peak at ∼ 540–570K and HD maxima at 540–590, 700–730 and 900–930K. It is suggested that deuterium is mostly retained in the bulk of steel material on interfaces of carbide precipitates and on grain boundaries.

Beryllium-tungsten study on mixed layers obtained by m-HiPIMS/DCMS techniques in a deuterium and nitrogen reactive gas mixture

Beryllium-tungsten 2μm thick layers were co-deposited on molybdenum, graphite and silicon substrates in a reactive argon-deuterium/argon-deuterium-nitrogen gas mixture by means of combining multi-pulse High Power Impulse Magnetron Sputtering and Direct Current Magnetron Sputtering techniques. The elemental composition and thickness of the layers were well controlled during the deposition process. Structural and chemical investigations were undertaken by means of X-ray diffraction and Rutherford backscattering methods. The results showed a change in the samples structure with the adding of nitrogen into the sputtering gas mixture, evidenced by the appearance of W2N peaks in the diffraction patterns. The RBS showed in-depth uniformity of beryllium and tungsten concentrations for the deposited coatings. In order to investigate the deuterium retention mechanisms and to make an assessment of the deuterium inventory, thermal desorption spectrometry was performed. The deuterium desorption spectra for the batch of samples obtained with nitrogen showed an obvious behaviour of shifting to higher temperatures. This behaviour may be related to the nitrogen desorption.

Deuterium retention in TiC and TaC doped tungsten at high temperatures

Samples made of tungsten doped either with titanium carbide (W–1.1TiC) or tantalum carbide (W–3.3TaC) were either exposed to D2 gas at a pressure of 100kPa at 800–963K or irradiated by 38eV/D ions at 800K. The deuterium (D) inventory in the samples was examined by nuclear reaction analysis and thermal desorption spectroscopy. The D bulk concentration and total retention in W–3.3TaC were comparable in all cases to that in pure polycrystalline W. The D bulk concentration in W–1.1TiC was more than one order of magnitude higher than that in pure W after exposure to D2 gas, and was also several times higher than that in W–1.1TiC after irradiation at 800K. It is suggested that D trapping inside the carbide precipitates in W–1.1TiC becomes essential at high temperatures.

Hydrogen trapping in carbon film: From laboratories studies to tokamak applications

During the DITS (Deuterium Inventory in Tore Supra) project, inner vessel walls of Tore Supra tokamak have been loaded with deuterium during a specific plasma experiment based on long pulse operation. Then, and in order to determine where the deuterium used as a plasma fuel was lost in the plasma facing components, the Tore Supra limiter had been dismounted and post mortem analysis had been performed on samples coming from different places corresponding to erosion or deposited area. In this the thermo-desorption spectrometry (TDS) studies are presented. All the TDS spectra present independently to their origin the same behaviour e.g. 5 desorption bands at 560 K, 820 K, 1050 K, 1238 K and 1375 K. These bands correspond to different trapping processes in carbon material which are identified with the help of laboratories experiments. Finally, the spectra are fitted by a rate equation model in which one of the main parameter which is the exponential pre-factor is experimentally evaluated (2.7·1015 s−1). This gives the possibility to unambiguously determine the trapping energies for the 5 trapping processes: 1.8 eV, 2.8 eV 3.6 eV, 4.2 eV and 4.7 eV. The model uses a dispersion in energy around these values due to the complex structure of Carbon Fibre Composite.

Dynamic plasma-wall modeling of ELMy H-mode with UEDGE-MB-W

The results of time-dependent coupled simulations of the boundary plasma and the first wall deuterium inventory in various sequences of many type-I ELMs with UEDGE-MB-W code are presented. The temporal evolution of deuterium inventories of the pedestal plasma and wall and the calculated rates of particle deposition into the wall during ELMs and of wall outgassing between ELMs are compared to the experimental data on DIII-D. The fraction of pedestal particle losses deposited into the wall during ELMs is studied for different sizes and frequencies of ELMs. Modeling results for the discharge exhibiting the transition from small to giant type-I ELMs due to NBI heating decrease are discussed, demonstrating the important role of wall outgassing in the pedestal density built-up. The dynamic deposition/release equilibrium attained in the saturated wall in a sequence of ELMs and the roles of different plasma-material interaction processes in generating gas release are analyzed.

Deuterium retention in TiC and TaC doped tungsten under low-energy ion irradiation

Deuterium retention in toughened, fine-grained, recrystallized tungsten-based materials doped either with titanium carbide or tantalum carbide was investigated. Samples were either irradiated by 200 eV D−1 ions to fluences in the range 1 × 1022–1 × 1024 D m−2 at a temperature of about 310 K or by 38 eV D−1 ions to fluences in the range 6 × 1022–6 × 1024 D m−2 at temperatures of 300 and 600 K. The deuterium inventory in the samples was examined by nuclear reaction analysis and thermal desorption spectroscopy (TDS). At low irradiation temperatures, deuterium retention in both doped materials was comparable to retention in pure tungsten (both recrystallized and as-received), while the shapes of TDS spectra and their evolution with the fluence were different. Deuterium retention in both doped materials at 600 K and at the highest fluence was higher than that in both recrystallized and as-received pure tungsten, and this can be attributed to the carbide precipitates.

Electron density evolution after L–H transitions and the L–H/H–L cycle in ASDEX Upgrade

The development of the electron density profile and its dependences after the L–H transition have been investigated. Only electron cyclotron resonance heated H-modes have been analysed to exclude core particle fuelling. While the density gradient in the edge transport barrier increases significantly after the L–H transition, the pedestal top temperature rises continuously with the applied heating power and shows no pronounced change at the transition. The H-mode density saturates at a level which correlates with the neutral gas density in the divertor prior to the L–H transition. Although the density build-up varies with the available deuterium inventory, the initial increase in the edge density gradient is similar. This has been observed independent of the L-mode plasma collisionality (ν* ≈ 2.8–5.5). The analysis of electron density and temperature profiles reveals that L–H and H–L transitions occur at similar pedestal top pressures, but the pedestal top densities are always higher at the time of the H–L back transition.

Structural analysis of eroded carbon fiber composite tiles of Tore Supra: insights on ion transport and erosion parameters

As part of the deuterium inventory in the Tore Supra project, a detailed structural analysis of the eroded tiles (carbon fiber composite tiles) of the main plasma-facing component of Tore Supra, the toroidal pump limiter, has been performed. It combines scanning electron microscopy, atomic force microscopy and Raman microscopy. A tile cross-section has been scanned, showing the eroded profile of the tile and allowing the net erosion rate to be determined. Ripples are observed on the surface of the eroded tiles and this pattern gives a clear indication of the ion flux direction. The height profile along the valleys indicates that the fibers are eroded faster than the matrix, the difference in the erosion rates probably originating from differences in microstructure.

X-ray micro-tomography studies on carbon based composite materials for porosity network characterization

Quantitative characterization of the porosity structure of the carbon reinforced carbon fibre (CFC) materials is carried out by high resolution cone-beam X-ray micro-tomography (CBμCT). It is shown that CBμCT provides useful information pertaining to the in-depth fuel migration into carbon tiles function of the CFC material structure. The investigated materials comprised two types of non-irradiated CFC samples (former ITER reference CFC NB31 and JET CFC DMS780) and a series of CFC N11 samples in the frame of the Deuterium Inventory in Tore Supra (DITS) post-mortem analysis. A procedure for the quantitative evaluation of the CFC porosity factor has been introduced and tested. Useful information about the pores connectivity as well as the metal impregnation inside the CFC macroscopic pores, in case of heat sink region of the TS CFC, can be retrieved. The method can be used for the quality control monitoring of the new CFC ITER reference materials.

Deuterium Inventory in Tore Supra (DITS): 2nd post-mortem analysis campaign and fuel retention in the gaps

A dedicated study on fuel retention has been launched in Tore Supra, which includes a D wall-loading campaign and the dismantling of the main limiter (Deuterium Inventory in Tore Supra, DITS project). This paper presents new results from a second post-mortem analysis campaign on 40 tiles with special emphasis on the D retention in the gaps. SIMS analysis reveals that only 1/3 of the thickness of deposits in the plasma shadowed zones are due to the DITS wall-loading campaign. As pre-DITS deposits contain less D than DITS deposits, the contribution of DITS to the D inventory is about 30–50%. The new estimate for the total amount of D retained in the Tore Supra limiter is 1.7×1024atoms, close to the previous estimate, with the gap surfaces contributing about 33%. NRA measurements show a stepped decrease of D along the gap with strong asymmetries between different gap orientations.

Influence of the microstructure on the deuterium retention in tungsten

The microstructure of tungsten samples is systematically modified by recrystallization to investigate the structure dependence of deuterium (D) retention. After a thorough characterization by scanning and transmission electron microscopy, non-recrystallized samples and samples recrystallized by different degrees are loaded with deuterium in a low-temperature plasma device. The deuterium inventory is measured by nuclear reaction analysis and thermal desorption spectroscopy. The post implantation surface morphology is investigated by scanning electron, optical and atomic force microscopy. The modification by recrystallization allows a wide variation in the crystallite size and has a strong impact on the measured D retention. Both the total amount and the binding state of the retained D are changed.

Deuterium inventory in the full-tungsten divertor of ASDEX Upgrade

The deuterium inventory in tungsten-coated divertor tiles used during the first full-tungsten plasma-facing wall phase of ASDEX Upgrade was measured by various methods of analysis. The D inventory in the inner divertor was still dominated by codeposition with residual carbon, whereas it was dominated by trapping in the thicker vacuum plasma sprayed tungsten layers at the outer divertor. The total inventory in the divertor area decreased by a factor of 5–10 compared with the period of carbon-dominated plasma-facing wall.

Deuterium Inventory in Tore Supra: status of post-mortem analyses

A dedicated study on fuel retention has been launched in Tore Supra which included a D wall-loading campaign and the dismantling of the main limiter (Deuterium Inventory in Tore Supra, DITS project). This paper presents new results of post-mortem analyses performed on selected limiter tiles. Scanning electron microscopy shows thick layers with poloidally oriented tip-shaped structures in deposition zones. In the erosion zone deposits inside the open porosities of the bulk material of the tiles were found. Raman microscopy indicates that hard deuterated amorphous carbon layers are deposited on the limiter tiles and on their gap sides. Secondary ion mass spectrometry D depth profiles show an increase of the D content near the bottom of the gap, in agreement with previous results. The campaign markers (13C and 11B) are found in only 2 out of 7 measurements in the deposited layers. Finally, the agreement of the SIMS data with previous nuclear reaction analysis (NRA) measurements is discussed, especially discrepancies possibly due to the non-uniformity of the samples.

Deuterium retention in different tungsten grades

Deuterium retention in different tungsten (W) grades was investigated for various incident ion energies ranging from 3 to 200 eV per deuterium atom and fluences ranging from 1×1023 m−2 to 2×1025 m−2. Irradiation temperatures were 320 and 500 K. The retained amount was determined by ion beam analysis and thermal desorption spectroscopy. For irradiation at 320 K, deuterium retention in plasma-sprayed tungsten (PSW) is higher compared to polycrystalline W for all investigated energies. For irradiation at 500 K, the deuterium retention for polycrystalline W increases strongly with fluence in contrast to PSW. The highest deuterium inventory was found for ITER grade W irradiated with 200 eV per deuterium at 500 K.

Dynamic and static deuterium inventory in ASDEX Upgrade with tungsten first wall

Hydrogen retention in the divertor tokamak ASDEX Upgrade is studied by surface analysis and gas balances methods. Comparing carbon and tungsten plasma facing components (PFCs), the deuterium content of deposits at the divertor plates has dropped by a factor of 13. With tungsten PFCs only 0.7% of the puffed hydrogen is retained, including a significant amount deeply implanted in the tungsten coatings at the outer divertor. Gas balances for ITER relevant high density H-mode discharges lead to a hydrogen retention averaged over a discharge of 8.2 ± 3.3% for tungsten PFCs and 23 ± 7% for carbon PFCs. For tungsten PFCs wall saturation is observed, i.e. only 1.5 ± 3.5% of the puffed gas is retained after reaching steady state conditions. Within the error bars all hydrogen outgasses within 15 min.