Tritium Retention Research Articles

Manufacture of thick VPS W coatings on relatively large CuZrCr substrate and its steady high heat load performance

W material is considered as one of potential Plasma Facing Materials (PFMs) for its high melting point, excellent stability at elevated temperature, good thermal conductivity, excellent anti-plasma sputtering and low Tritium retention. Functionally graded W/Cu coating was applied on CuCrZr substrate (250mm×120mm×30mm) with compositionally gradient W/Cu as bond coat (0.4–0.6mm) and 1.5mm thick W coating as top coat via Vacuum Plasma Spraying (VPS) for continuous deposition of 5h. Microstructure, chemical composition, porosity and adhesive strength for as sprayed thick W coating on the CuCrZr substrate were characterized by means of SEM, ICP-MS, Mercury Intrusion Porosimeter and tensile strength tester. The steady high heat load (HHL) performance for W/Cu functional gradient coating was evaluated by high energy electron beam. The results showed that thick VPS W coated CuCrZr substrate can withstand the steady high heat load at the electron beam power density of 9MW/m2 for 1000cycles.

Tritium retention properties of tungsten, graphite and co-deposited carbon film

DT+ ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both the amount of retained tritium and the reduction of retained tritium after preservation in vacuum were investigated using an IP technique and BIXS. In addition, the relationship between the retention properties of tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. The amount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1keV of DT+ irradiation, graphite showed no reduction of the amount of retained tritium after six months preservation while that of carbon film decreased by approximately 20% after 40 days preservation. It was suggested that this difference might be associated with differences in the microstructure between graphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately half after 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decrease in tritium retention during preservation became slower. Tungsten's properties of releasing tritium while preserved in vacuum would be a useful tool for the reduction/removal of retained tritium.

Estimation of the contribution of gaps to tritium retention in the divertor of ITER

An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s−1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas.

Hydrogen transport in solids with traps in the case of continuum distribution of detrapping energies

Tritium retention in the first wall material is one of the key issues in the performance of future fusion reactors. Transport of hydrogenic species in these materials is most commonly treated as diffusion affected by trapping/detrapping processes. Usually only several trap types differing in their activation energies of hydrogen release are considered (up to three types in the TMAP7 code). We suggest that in some cases (e.g. highly damaged or disordered media) the hydrogen trapping/detrapping process is better characterized by a continuum distribution of traps over their detrapping energies. Within a random walk model we show that this assumption leads to qualitative changes in hydrogen transport in solids. Using this model we explain experimental findings on temporal dependence of deuterium outgassing from tokamaks, first wall.

In situ deuterium inventory measurements of a-C:D layers on tungsten in TEXTOR by laser induced ablation spectroscopy

Laser induced ablation spectroscopy (LIAS) is a diagnostic to provide temporally and spatially resolved in situ measurements of tritium retention and material migration in order to characterize the status of the first wall in future fusion devices. In LIAS, a ns-laser pulse ablates the first nanometres of the first wall plasma-facing components into the plasma edge. The resulting line radiation by plasma excitation is observed by spectroscopy. In the case of the full ionizing plasma and with knowledge of appropriate photon efficiencies for the corresponding line emission the amount of ablated material can be measured in situ. We present the photon efficiency for the deuterium Balmer α-line resulting from ablation in TEXTOR by performing LIAS on amorphous hydrocarbon (a-C:D) layers deposited on tungsten substrate of thicknesses between 0.1 and 1.1 μm. An experimental inverse photon efficiency of was determined. This value is a factor 5 larger than predicted values from the ADAS database for atomic injection of deuterium under TEXTOR plasma edge conditions and about twice as high, assuming normal wall recycling and release of molecular deuterium and break-up of D2 via the molecular ion which is usually observed at the high temperature tokamak edge (Te > 30 eV).

Thermal FEA for Alcator C-Mod Advanced Outer Divertor

An advanced outer divertor is being developed for Alcator C-Mod to study reactor fuel (tritium) retention and plasma wall material interaction physics at reactor temperatures with high power long-pulse discharges. The divertor will be operated at controlled temperature of 600 °C. To achieve this goal, the divertor will be structurally and electrically continuous along the toroidal direction, requiring it to expand radially as temperature increases. This paper describes the thermal finite element analysis (FEA) and results of the outer divertor. There are four aspects, with focus on the A-Frame assembly. First of all, a one twentieth module of the full divertor is composed of divertor tiles, tile mounting plate, heaters, divertor gusset, A-Frame support, spherical bearings, bracket, halo current shunt, vessel gusset, and so on. By adjusting the power of each of the seven toroidal divertor heaters, the tiles achieve a uniform temperature poloidally with toroidal temperature variation within allowables. The temperature of each component is evaluated, and results are used to support the design changes. Second, radiation simulation on multilayer radiation shields behind divertor plate is presented. Third, radiation simulation of the diverter heater itself is done to understand more details of heat transfer from the heater to the surrounding tiles and support plates. Finally, thermal analysis is completed with a model including a tile and its mounting plate, to predict the effect of plasma heat load on divertor tiles. All the thermal FEA was performed with COMSOL, a commercial FEA software.

Blanket material and technology developments toward DEMO under the Broader Approach framework

In the Broader Approach (BA) framework, research and development on blanket related materials and technologies have been carried out between the EU and Japan. Those activities are implemented mainly at the Rokkasho BA site in collaboration with universities in Japan. In the R&D on SiCf/SiC composites for an advanced blanket material, mechanical properties of chemically vapor infiltrated (CVI) SiCf/SiC composites have been obtained in high temperature vacuum environment up to 1000°C. As to R&D on the tritium technology, tritium retention of the fine-grained re-crystallized tungsten has been evaluated. On reduced activation of ferritic/martensitic (RAFM) steels as the blanket structural material, a 20-ton heat of the F82H RAFM steel has been successfully produced in an electric arc furnace. Advanced neutron multiplier pebbles of beryllide (beryllium–titanium alloy) have been fabricated using a dedicated rotating electrode apparatus followed by annealing from the beryllide rod produced with the plasma sintering method. Also advanced tritium breeder pebbles of lithium-rich lithium titanate have been fabricated by an emulsion method, where the grain size, confirmed by SEM, was less than 5μm. Tritium recovery tests of advanced tritium breeder pebbles will be investigated under 14MeV neutron irradiation as one of the new activities.

Tritium Retention on the Surface of Stainless Steel Samples Fixed on the Plasma-Facing Wall in LHD

Tritium Retention on the Surface of Stainless Steel Samples Fixed on the Plasma-Facing Wall in LHD

Impurity seeding for tokamak power exhaust: from present devices via ITER to DEMO

A future fusion reactor is expected to have all-metal plasma facing materials (PFMs) to ensure low erosion rates, low tritium retention and stability against high neutron fluences. As a consequence, intrinsic radiation losses in the plasma edge and divertor are low in comparison to devices with carbon PFMs. To avoid localized overheating in the divertor, intrinsic low-Z and medium-Z impurities have to be inserted into the plasma to convert a major part of the power flux into radiation and to facilitate partial divertor detachment. For burning plasma conditions in ITER, which operates not far above the L–H threshold power, a high divertor radiation level will be mandatory to avoid thermal overload of divertor components. Moreover, in a prototype reactor, DEMO, a high main plasma radiation level will be required in addition for dissipation of the much higher alpha heating power. For divertor plasma conditions in present day tokamaks and in ITER, nitrogen appears most suitable regarding its radiative characteristics. If elevated main chamber radiation is desired as well, argon is the best candidate for the simultaneous enhancement of core and divertor radiation, provided sufficient divertor compression can be obtained. The parameter Psep/R, the power flux through the separatrix normalized by the major radius, is suggested as a suitable scaling (for a given electron density) for the extrapolation of present day divertor conditions to larger devices. The scaling for main chamber radiation from small to large devices has a higher, more favourable dependence of about Prad,main/R2. Krypton provides the smallest fuel dilution for DEMO conditions, but has a more centrally peaked radiation profile compared to argon. For investigation of the different effects of main chamber and divertor radiation and for optimization of their distribution, a double radiative feedback system has been implemented in ASDEX Upgrade (AUG). About half the ITER/DEMO values of Psep/R have been achieved so far, and close to DEMO values of Prad,main/R2, albeit at lower Psep/R. Further increase of this parameter may be achieved by increasing the neutral pressure or improving the divertor geometry.

On the reduction of deuterium retention in damaged Re-doped W

Modelling tritium retention in radiation-damaged pure-tungsten samples, the concentration of deuterium retained in the tungsten-ion-induced damage zone decreases with increasing exposure temperature; whereas for a similarly treated tungsten–rhenium alloy, it can drop significantly faster than in pure tungsten, given sufficiently high temperatures. In contrast, similar changes in retention behaviour were not observed in undamaged samples. Based on these findings, as well as on corresponding previous TEM results, it is concluded that the concentration of high-energy radiation-induced defects responsible for trapping of deuterium is lower in the alloy than in pure tungsten. Therefore, a small amount of transmutation rhenium in damaged tungsten should be able to keep tritium retention to a low level in ITER.

Hydrogen permeability through beryllium films and the impact of surface oxides

Beryllium will constitute the major part of the first wall of ITER, however, several aspects of the tritium retention and recycling in fusion reactors are still open. Studying details of the hydrogen isotope interactions on Be films is in principle easier and more accurate than on the bulk Be metal since a thin (and therefore more permeable) layer of Be film could be deposited on a desired substrate by applying well controlled methods. Results of the hydrogen permeation through 8 micrometer thick Be films deposited by the thermionic vacuum arc method on Eurofer steel membranes with exposed area of 8.4cm2 are presented. The permeation reduction factor (PRF) at 400°C varied on six samples from 14 to 135 with respect to the bare Eurofer membrane. The highest PRF value enables expression of the Be film permeability coefficient P by means of a simple model which gives PBe∼2×10−15molH2/msPa0.5. Lower PRF values could be explained by microscopic imperfections which represent parallel hydrogen paths through the Be film and enhance the permeation rate. Some of them were revealed by the SEM while their presence could be confirmed also by observing permeation flux transients recorded after the hydrogen exposure. The two-step process of achieving the steady flux agrees with our numerical simulation. It was found that for unintentionally oxidized samples the extracted regular (eliminated contribution of the pinholes in Be film) permeation rate is almost identical from sample to sample and accounts to j≈1.2×10−7H2/m2 s at 1bar hydrogen driving pressure due to BeO formation. For a non-oxidized sample this value is several times higher, j≈6.5×10−7molH2/m2s. From the latter follows that PBe≈1.9×10−14molH2/msPa0.5, while PBeO∼1×10−17molH2/msPa0.5 can be estimated by assuming a 35nm thick BeO layer.

Application of laser-induced breakdown spectroscopy for characterization of material deposits and tritium retention in fusion devices

Laser-induced breakdown spectroscopy (LIBS) is discussed as a possible method to characterize the composition, tritium retention and amount of material deposits on the first wall of fusion devices. The principle of the technique is the ablation of the co-deposited layer by a laser pulse with P (power density)≥0.5GW/cm2 and the spectroscopic analysis of the light emitted by the laser induced plasma. The typical spatial extension of the laser plasma plume is in the order of 1cm with typical plasma parameters of ne≈3×1022m−3 and Te≈1–2eV averaged over the plasma lifetime which is below 1μs. In this study “ITER-Like” mixed deposits with a thickness of about 2μm and consisting of a mixture of W/Al/C and D on bulk tungsten substrates have been analyzed by LIBS to measure the composition and hydrogen isotopes content at different laser energies, ranging from about 2J/cm2 (0.3GW/cm2) to about 17J/cm2 (2.4GW/cm2) for 7ns laser pulses. It is found that the laser energies above about 7J/cm2 (1GW/cm2) are needed to achieve the full removal of the deposit layer and identify a clear interface between the deposit and the bulk tungsten substrate by applying 15–20 laser pulses while hydrogen isotopes decrease strongly after the first laser pulse. Under these conditions, the evolution of the spectral line intensities of W/Al/C/hydrogen can be used to evaluate the layer composition.

Hydrogen trapping in 3He-irradiated Fe

The characteristics of irradiation-induced hydrogen (deuterium) traps in pure Fe were investigated for quantitative evaluation of tritium retention in fusion reactor components. The deuterium depth profiles of an Fe disk sample exposed to deuterium plasma were observed by means of nuclear reaction analysis (NRA) before and after irradiation with 0.8MeV or 1.3MeV 3He ions. Irradiation generated a number of traps, and deuterium retention was drastically increased subsequent to irradiation. Steady-state deuterium concentration in the trap and the solution sites were obtained by continuously charging the sample with deuterium during the NRA. Based on these values, the trapping energy, which is the enthalpy difference between the two sites, was estimated to be 0.38eV. The number ratio of the trap to atomic displacement was 0.013. Some of the traps were annihilated around 523K. The annihilation temperature, the trapping energy, and the equilibrium constant suggest that the trap is a dislocation loop introduced by the irradiation. It is deduced that the tritium inventory in the Fe components of a reactor should be drastically increased by neutron irradiation due to the formation of traps, but may be significantly reduced by high temperature operation of the components.

Hydrogen Solubility of FLiNaK with Hydrogen Plasma Interaction

Hydrogen solubility of FLiNaK (LiF 46.5 mol% + NaF 11.5% + KF 42 mol%) were investigated with hydrogen plasma interaction. To use molten salt as liquid wall for fusion device, tritium retention property of the molten salt should be studied. Although there have been some reports on hydrogen solubility of FLiNaK, retention property of FLiNaK with hydrogen plasma interaction has not been reported yet. Hydrogen outgassing of molten FLiNaK was measured with RGA after interaction of hydrogen plasma. Hydrogen partial pressure of the RGA was calibrated with an H2 mass flow controller of 5sccm. Hydrogen plasma was generated with 500W ECR source and the molten FLiNaK was contained with heated crucible of diameter of 46mm and depth of 40mm. By measuring hydrogen ion density near the surface of FLiNaK, we evaluated dose of hydrogen to the FLiNaK and calculated retention percentage with measured outgassing amount of hydrogen. Compared to Henry’s law, plasma-interacted-FLiNaK showed significantly large amount of hydrogen retention such as 10000 times.

Hydrogen traps in ion-irradiated F82H steel observed by NRA

Characteristics of irradiation-induced trapping of hydrogen were investigated for quantitative evaluation of tritium retention in F82H steel. Before and after irradiation of 0.8-MeV 4He or 0.3-MeV H ions, deuterium depth profiles near the surface of a disk sample were observed by nuclear reaction analysis under continuous exposure of deuterium plasma. One type of trap, with a trapping energy of 0.66eV, was observed after each irradiation. The ratio of trap production rate to atomic displacement was 0.0046 and 0.0014 for He and H irradiation, respectively. Annihilation occurred around 600K for H irradiation but was not observed even at 691K for He irradiation. Traps are likely to be interstitial-like sites associated with dislocation loops. This study also indicates that helium plays a role in inhibiting trap annihilation. In addition, the deuterium diffusion coefficient in non-irradiated F82H was determined by a time-lag permeation experiment.

Formation of lithium-tritide by hot atom reactions of tritium produced in Pb-16Li

The release behaviors of hydrogen isotopes in Pb-16Li introduced by thermal gas exposure or produced by thermal neutron irradiation were compared to investigate hot atom reactions of tritium. The solubility of hydrogen isotope was evaluated to be S=6.56×10−7 exp(−0.11 [eV]/kT) [at. fr, Pa0.5], which is consistent with the literature values. The tritium TDS spectra for Pb-16Li with thermal neutron irradiation showed that about 5% of tritium was released at single release stage around 600K, which was higher than the melting point of Pb-16Li, although no release peak was found at around 600K for the hydrogen isotope-doped Pb-16Li. The kinetic analysis indicated that the tritium release would be associated with the dissociation of Li-T bond, indicating that LiT was formed in Pb-16Li by thermal neutron irradiation, suggesting that the formation of Li-T bonds should be considered to estimate tritium retention in Pb-16Li eutectic blanket systems in fusion reactors.

Tritium retention to the first wall of JT-60U

For economic and safety reasons, tritium (T) accumulation on plasma facing wall (PFW) of fusion reactor is strictly limited. In this study, T inventory in the graphite tiles used at the first wall of JT-60U was measured by a full combustion method. It was found that T was only retained near plasma facing surfaces sides of the tiles and the amount of retained T increased from <1011 to <1013T atoms/cm2 with increasing the exposed discharge period of the tiles. Integrating the T retention with the total surface area of the outer first wall tiles, the fraction of retained T in that area was estimated to be 13% of the total T production. It was confirmed that these retained T were part of the energetic T produced by DD reactions without being replaced by HH discharges. Based on the retained fraction, the annual amount of T retention in the outer first wall of a demo-size reactor was calculated to be 360g/burn-year at maximum. Even the value would be much less in the reactor, the accumulation of this kind of inventory could have significant contribution to the total T inventory.

Determination of hydrogen diffusion and permeation coefficients in pure copper at near room temperature by means of tritium tracer techniques

• We have applied a tritium tracer technique for gaseous hydrogen permeation in Cu. • We have succeeded to get reliable data for hydrogen permeability in Cu. • Diffusivity are bending downward from the extrapolation of higher temperature. • Diffusivity are influenced by initial surface contamination which is removed by hydrogen. Copper (Cu) and its alloys are candidate materials for heat sinks or cooling-tubes in a fusion reactor. Hence their tritium retention and permeation are very important safety concerns. Most data for diffusion and permeation of hydrogen in Cu so far available have been limited for rather higher temperatures and data for lower temperatures, in particular, for near room temperature (RT) are scarce. We have applied a tritium tracer technique for gaseous hydrogen permeation in pure Cu at near RT and succeeded to get reliable data for hydrogen permeation coefficients given by Φ = (2.8 ± 0.4) × 10 −6 exp(−85 ± 2(kJ/mol)/RT), mol m −1 s −1 Pa −1/2 , which is reliable in very wide temperature range from 300 K to 1000 K. However, diffusion coefficients determined by the time-lag method are bending downward from the extrapolation of higher temperature data and are influenced by initial surface contamination which is removed by hydrogen loading.

Characterization of Tritium Retention in Plasma Sprayed B4C/Mo Coatings

Quantification of the amount and depth of tritium retained on plasma facing materials(PFMs) surface is of great importance for selecting PFMs with high safety in the nuclear system. In the present study, potential PFMs of B4 C/Mo coatings were prepared and the depth profiles of tritium in the coatings were measured both by imaging plate(IP) technique and β-ray induced X-ray spectrometry(BIXS). The IP images showed that the amount of adsorbed tritium in the coatings followed the order of B4 C BM15 BM5 Mo, which was along with the X-ray spectra also illustrating that most tritium penetrated into the bulk of the B4 C coating while on the surface only for the other three coatings. The cross-sectional morphologies of the coatings were observed by scanning electron microscope(SEM), which demonstrated that B4 C coatings had the highest porosity while the other three coatings were denser despite of the existence of smaller-sized pores and cracks. All the results suggest that pores and microcracks in the as-sprayed coatings provide passageway for tritium penetration and adsorption, which are the key factors that affect the amount of tritium adsorbed. Furthermore, the content of Ti contamination in the coatings also plays an important role in controlling tritium adsorption.

Tritium release and retention properties of highly neutron-irradiated beryllium pebbles from HIDOBE-01 experiment

The current helium cooled pebble bed (HCPB) tritium breeding blanket concept for fusion reactors includes a bed of 1mm diameter beryllium pebbles to act as a neutron multiplier. Beryllium pebbles, fabricated by the rotating electrode method, were neutron irradiated in the HFR in Petten within the HIDOBE-01 experiment. This study presents tritium release and retention properties and data on microstructure evolution of beryllium pebbles irradiated at 630, 740, 873, 948K up to a damage dose of 18dpa, corresponding to a helium accumulation of about 3000appm. The measured cumulative released activity from the beryllium pebbles irradiated at 948K was found to be significantly lower than the calculated value. After irradiation at 873 and 948K scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses revealed large pores or bubbles in the bulk and oxide films with a thickness of up to 8μm at the surface of the beryllium pebbles. The radiation-enhanced diffusion of tritium and the formation of open porosity networks accelerate the tritium release from the beryllium pebbles during the high-flux neutron irradiation.