Tritium Desorption Research Articles

The diffusion and desorption of tritium on Li2TiO3 (-133) surface from first-principles calculations

Li2TiO3 is one of the most promising solid breeder materials for nuclear fusion reactor. Diffusion on the surface is a vital step of the tritium release process. In this work, the diffusion behavior of tritium atoms on the (-133) surface and the desorption of T2 molecules from the (-133) surface are systematically investigated. Possible adsorption sites, local diffusion pathways, and their corresponding activation energies have been identified using first-principles calculations and the climbing image nudged elastic band (CI-NEB) method. The diffusivity and effective activation energy for tritium diffusion on the (-133) surface have been determined using kinetic Monte Carlo (KMC) simulations. The activation energy for T2 desorption from the (-133) surface has also been investigated. The effective activation energy for tritium diffusion on the (-133) surface is found to be 1.06 eV, whereas the activation energy for T2 molecule desorption is 1.46 eV. The (-133) surface exhibits a higher activation energy for tritium diffusion but a lower activation energy for T2 desorption compared to the (001) surface.

Dimensional effects in analysis of laser-induced-desorption diagnostics data

The accurate assessment of the local tritium concentration in the tokamak first wall by means of the laser-induced desorption (LID) diagnostics is sought as one the key solutions to monitoring the local radioactive tritium content in the first wall of the fusion reactor ITER. Numerical models of gas desorption from solids used for LID simulation are usually closed with the one-dimensional transport models. In this study, the temperature and particle dynamics in the target irradiated by a short laser pulse during LID are analyzed by means of the two-dimensional model to assess the validity of using one-dimensional approximation for recovering the diagnostics signal. The quantitative estimates for the parameters governing the heat and particle transfer are presented. The analytical expressions for the sample spatiotemporal temperature profiles driven by the target irradiation with a Gaussian laser beam with the trapezoid temporal shape are derived. The obtained relations are used to simulate tritium desorption from a tungsten sample driven by pulsed heating. It is shown that depending on the ratio between the laser spot radius and the heat diffusion length, the one-dimensional approach can noticeably overestimate the sample temperature in the limit of small laser spot radius (estimated for tungsten as ∼0.5–1.0 mm), resulting in more than 100% larger amounts of tritium desorbed from the target, compared to the two-dimensional approximation. In the limit of large laser spot radius (≥1.5 mm), both approaches yield comparable amounts of tritium desorbed from the sample.

Correction: Chakin et al. Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor. J. Nucl. Eng. 2023, 4, 552–564

The authors would like to make the following corrections to the published paper [...]

Comparative analysis of gas release from biphasic lithium ceramics pebble beds of various pebbles sizes and content under neutron irradiation conditions

This paper presents the results of 4 reactor campaigns on the irradiation of biphasic lithium ceramics containing different ratios of lithium orthosilicate (LOS) and lithium metatitanate (LMT) components (25 and 35 mol% LMT in LOS). The size distribution of pebbles in pebble beds was 250–1250 μm and 500–710 μm, respectively. The studies were carried out sequentially with each type of ceramics. In experiments carried out using the vacuum extraction method, the composition of gases released from lithium ceramic samples was registered in in-situ mode. The absence of purge gas during the experiments minimized the possibility of T2O and HTO formation, significantly simplifying processing of the results and providing more opportunities for results analysis. The main goal of the present paper was to identify the general patterns of the release of gases with mass numbers M2 (H2), M4 (He + HT), M6 (T2) and M18 (H2O) from ceramic samples throughout the entire irradiation experiment in 4 campaigns. Release trends of main gases with mass numbers M2, M4, M6 and M18 for all four campaigns are presented and their comparative analysis was performed. The average partial pressure of tritium release in the form of T2 and HT molecules for all campaigns was determined as 5.8 × 10-7 Torr. The dependences of formation rates of helium release peaks on the irradiation time were plotted. The nature of peak emissions does not have a monotonic relationship; upon irradiation, both an increase in the frequency of peaks and a decrease in it are observed. During irradiation, the process of peak helium release does not stop.The simulation was carried out by the finite element method, assuming that tritium release from the sample is determined by diffusion and desorption processes from the sample surface. The experimental curves are satisfactorily described by a number of sets of desorption and diffusion parameters. One way or another they lie in the range of specified values. The Arrhenius dependences of the effective diffusion coefficient and desorption coefficient obtained for lithium ceramics 35 LMT are equal to:D=5,2×10-11m2sexp-21kJmoleRT,K=1.21×10-4m2sexp-64kJmoleRT.The values of the effective diffusion coefficient and tritium desorption coefficient in 25 LMT ceramics were 15 and 20% lower than in 35 LMT ceramics.

Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor

The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.

Thermal desorption of tritium and helium from lithium ceramics Li2TiO3+5mol% TiO2 after neutron irradiation

Lithium-based ceramics exhibit good thermophysical properties, have low activation and chemical activity, and also release tritium well. These characteristics make lithium ceramics the best candidates for use as a functional material for a solid breeder blanket. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials can only be obtained from the results of reactor experiments. On the other hand, many important parameters can only be estimated in post-irradiation experiments (PIE). This paper describes studies of tritium and helium release in post-irradiation experiments (PIE) on the thermal desorption from ceramic pebbles Li2TiO3 + 5mol% TiO2 (with 96% enrichment on lithium-6 isotope), irradiated in the WWR-K reactor for 223 days. In PIE a peak of helium release was recorded for each sample in the temperature range of 1300–1500 K. An assumption was made in the paper to explain the nature of this helium release peaks from lithium ceramic pebbles.

Effect of neutron dose on the tritium release behavior of Li2TiO3–0.5Li4SiO4 biphasic ceramic

Neutron irradiation with different fluence and flux on the Li2TiO3-0.5Li4SiO4 biphasic ceramic powders was conducted. The tritium release behavior was studied by thermal desorption spectroscopy (TDS) with different heating rates and isothermal heating at various temperatures. It was found that the tritium TDS spectra were shifted toward the higher temperature side as the neutron fluence was reduced. The tritium migration in the biphasic ceramic was governed by diffusion and de-trapping processes with the activation energies of 0.29–0.50 eV and 1.00–1.13 eV for the neutron-irradiated samples with the fluence of 8.25 × 1015 and 3.96 × 1016 n cm−2. For isothermal heating, there is another tritium desorption peak when the sample was continually heated to higher temperatures after isothermally heating at 690 K and 800 K, indicating that tritium still remained in the ceramic under isothermal heating conditions. The intensity of the second peak was decreased at the higher isothermal heating temperatures. The diffusion model with detrapping reaction can fit the experimental data, suggesting that both the detrapping and diffusion processes should be considered to estimate tritium release characteristics for the tritium breeding biphasic ceramics.

Electrochemical Sensors and Techniques for Redox Potential and Tritium Transport in a Neutron-Irradiated Molten Flibe Salt Loop

A consortium of the Massachusetts Institute of Technology, North Carolina State University, the University of California at Berkeley (UCB) and Oak Ridge National Laboratory have initiated a project to build a neutron-irradiated molten-salt forced-circulation loop at the MIT reactor. The loop will use FLiBe (Li2BeF4) salt and will duplicate thermal-hydraulics, chemical, and neutronics conditions in a salt reactor.The SALT Lab at UCB is responsible for the design and application of electrochemical sensors and techniques for redox measurement and tritium transport within the loop. The proposed experiments will investigate redox measurements and control in the loop, tritium retention and diffusion in graphite, and tritium transport in the salt. Redox measurements and control studies will be performed to quantify the corrosive effect of neutron activation reactions and the effect of redox control agents in the loop. Tritium studies are being developed to quantify tritium uptake capacity in graphite, identify tritium desorption mechanisms, and measure tritium solubility, diffusivity, and speciation in the melt. To support these experiments, a variety of electrochemical probes are under development at UCB, including different size, shape, and material of the electrodes.This talk will present the electrochemical probes under development and discuss the experiments that will be tested off-loop in the Be-gloveboxes of the SALT lab and then implemented in the loop.

Simulation calculation of the influence of interstitial atoms on the desorption behavior of tritium in nuclear graphite

In order to explore the effect of interstitial atoms on tritium desorption behavior of nuclear graphite, the adsorption and desorption behaviors of hydrogen atoms on interstitial atoms were simulated by first-principles. The results indicated that the interstitial atoms can gradually adsorb 3 hydrogen atoms until the sp3 hybrid structure is formed. The CH3 structure with dangling bonds tends to desorb as ionic CH3, which will combine with free H or other CH3 to form CH4 or C2H6. The CH (or CH2) structure tends to polymerize with other CH (or CH2) structures to form C2H2 (or C2H4), which reaction energy barrier is slightly lower than that of the CH3 structure; Although the migration energy of the interstitial atoms with H adsorbed on the graphite flakes is not much different from that of the interstitial atoms without H, the energy barrier of their recombination with single vacancies is significantly greater than that of the interstitial atoms without H. Therefore, it can predict that the large number of interstitial atoms in decommissioned nuclear graphite will become important sites for chemisorption of tritium. In addition, the reaction energy barrier of the interstitial atoms possessed large adsorption energy with tritium gas desorption is significantly higher than that with carbon containing tritium products (including C2T2, C2T4, CT4 or C2T6). This work will provide technical guidance and support for the heat treatment of the adsorbed tritium in decommissioned nuclear graphite.

The effect of long-term heating on the tritium adsorption and desorption behavior of advanced core–shell breeding materials

In order to efficiently grasp the tritium behavior in advanced core–shell breeding materials, this study adopts a route of injecting tritium out-of-pile to deal with the problems of fewer platforms, long periods, and high costs for traditional neutron irradiation-tritium release experiments. Here, tritium adsorption and desorption experiments were carried out with Li4TiO4–Li2TiO3 core–shell breeding materials before and after long-term heating up to 30 days. The purpose is to study whether the structural changes caused by long-term heating will affect the adsorption and desorption behavior of tritium on the sample surface. The results show that the lack of chemically adsorbed water caused by long-term heating will significantly weaken the tritium adsorption capacity of the sample, but will not affect the desorption behavior of Ar, 1% H2 + Ar and water vapor on tritium, so all samples show a very low tritium retention.

Titanium beryllide as an alternative to beryllium in nuclear and thermonuclear engineering, capabilities of UMP JSC in the technology development and beryllides products manufacture

For many decades, beryllium has been used as a structural element in nuclear installations as a moderator / breeder of fast neutrons. The consequence of neutron irradiation is a significant production of gas products in the form of helium and tritium, which leads to swelling and loss of strength properties of beryllium reflectors. The relatively low melting point of beryllium also imposes restrictions on the high-limit temperature regimes of the reactor core. As an alternative to pure beryllium, it is necessary to consider intermetallic compounds based on it, in particular titanium beryllide. Preliminary studies on the thermal desorption of helium and tritium from titanium beryllide have shown that this material has a much lower retention tendency and a lower release temperature. The higher melting point of titanium beryllide compared to pure beryllium is also an advantageous characteristic.Over the past years, UMP JSC, thanks to its research in this area, has achieved significant success in the development of technology for obtaining intermetallic billets and articles based on titanium and chromium beryllides. As a technology demonstrator, prototypes of structural elements of a helium-cooled blanket breeder module of the projected DEMO reactor were made by order of the Karlsruhe Institute of Technology, Germany.The advantages of titanium beryllide, as well as the success achieved in the production of billets and products from it, open up opportunities for a more extensive study of the nuclear, physical and mechanical properties of this material with the possibility of further use in nuclear technology, including thermonuclear reactors, and in high-temperature instrumentation.

Modelling tritium adsorption and desorption from tungsten dust particles with a surface kinetic model

A kinetic surface model is presented and used to explain the loading and desorption kinetics of tritium retained in micrometre-sized tungsten (W) dust particles. The model describes the sticking of hydrogen isotopes from the gas phase to W surfaces and the desorption from W surfaces. The initial sticking coefficient is set to one and independent of the temperature. The activation energy for desorption depends on the hydrogen coverage of the surface and is parameterised with density functional theory (DFT) calculations for W(100), W(110) and W(111) surfaces. The DFT-parameterised model is successfully compared to experimental results showing that the amount of measured tritium as well as the desorption kinetic can be modelled with only tritium adsorbed on the surface of W dust particles. Then, the model is used to explore possible scenarios to remove the tritium from the W surfaces by exposing the tritiated surfaces to either deuterium and hydrogen. The simulations suggest that it can be possible to remove all the tritium trapped on the W surfaces even at room temperature as soon as the hydrogen or deuterium pressure is higher than the tritium pressure. This gives opportunity to build tritium removal scenarios for ITER.

Tritium adsorption and desorption on/from nuclear graphite edge by a first-principles study

The removal of tritium from irradiated nuclear graphite is one of the key targets of nuclear graphite decontamination. Here, we employ first principle density functional theory (DFT) to study its adsorption and molecular desorption on the most common edges and their reconstructions. The calculated adsorption energy ranges from −2.6 to - 5 eV, depending on the edge structure and the hydrogenation level, and are much larger than on the bulk of graphite. The hydrogenation level increases with the increase of hydrogen partial pressure, and drops rapidly at high temperature. The pathways to fully saturated edges are then determined for each edge variant and the activation energy (Eac) for molecular desorption computed by properly accounting for vibrational zero-point energy corrections (ΔEZPE). Our results showed that, there are three different stages for the desorption of hydrogen isotopes from nuclear graphite, namely stage 1 (200–300 °C), stage 2 (500–700 °C), stage 3 (1000–1100 °C). Our results can provide a theoretical basis for the tritium removal experiment from nuclear graphite.

Thermal Decontamination of Spent Activated Carbon Contaminated with Radiocarbon and Tritium

The thermal desorption of tritium (3H, T) and radiocarbon (14C) from spent activated carbon was investigated and three thermal desorption steps were established: the vaporization of homogeneously condensed molecules, the desorption of molecules physically binding with the carbon surface, and the decomposition of chemisorbed molecules. A model-free kinetic analysis was conducted to establish the optimum condition of vacuum thermal desorption. Physisorbed species, including tritiated water (HTO) and 14CO2, were effectively removed by vacuum thermal desorption. However, a fraction of 14C, which may take the form of carbon molecules in pyrocarbon form during the heating process, was not removed, even at a high temperature of 1000 °C under a vacuum of 0.3–0.5 Pa. Oxidative peeling of the pore surfaces by filling the evacuated pores with pure oxygen via vacuum thermal desorption and heating to 700 °C was found to be effective for reducing the level of 14C to a level below the established free-release criterion (1 Bq/g) when treating spent activated carbon with 14C radioactivity levels of 162 and 128 Bq/g. The reactivation of the spent granular activated carbon (GAC) by vacuum thermal desorption followed by surface oxidation was also confirmed by the slightly enhanced pore volumes when compared to those of virgin spent activated carbon.

Development of tritium dynamic transport analysis tool for tritium breeding blanket system using Modelica

Demonstration of the engineering feasibility of tritium breeding, including tritium generation, tritium extraction, tritium control and tritium safety is one of the main objectives of ITER Testing Blanket Module (TBM) Program. As one of TBM concepts, the tritium transport assessment of the China Helium Cooled Ceramic Breeder TBM and its ancillary systems (called test blanket system, TBS) is absolutely vital to understand and analyze the dynamic tritium behaviors in the system in detail according to the ITER operation scenario. Considering the requirements of the tritium transport analysis for the HCCB TBS, an integrated dynamic tritium transport analysis tool has been developed by the system simulation language “Modelica”. The multiple dynamic physics effects are considered in the tool, including the aspects impacting the tritium transport, simplified thermal-hydraulics considering only the linear loss and local loss in fluid, 1-d thermal conduction in solid, hydrogen isotope transport in fluid, 1-d tritium permeation in solid, tritium solution and recombination on solid surfaces, tritium desorption from ceramic breeder etc. A simplified model of the HCCB TBS is modeled to verify the functions and accuracy of this tool. The results, such as dynamic change of tritium concentration/partial pressure in fluids, dynamic change of tritium inventory in solids, tritium permeation fluxes etc., have been obtained, which are preliminary verified at various aspects by comparing with analytical and hand calculations.

Deuterium removal from beryllium co-deposits by simulated strike-point sweeping

Abstract Beryllium (Be) co-deposition with hydrogen isotopes will probably be the main mechanism of tritium accumulation in ITER. To reduce the use of baking of the first wall and divertor, raising strike-points during a pure deuterium (D) discharge has been proposed to heat the co-deposits directly and enhance tritium desorption. The feasibility of such plasma-induced removal as a potential means of tritium control has been investigated. Be-D co-deposited layers with a thickness of around 3 μm were created at 428–458 K and exposed to D or H plasma at various temperatures (673–873 K) for various times (180–1200 s). The remaining D content was measured by thermal desorption spectroscopy and compared to the control samples. Plasma-induced desorption was found to be efficient in removing significant amounts of D out of the co-deposits, removing 85% of D from the layer after D plasma exposure at 773 K for 1200 s and 84% of D after exposure at 873 K for only 180 s.

Performance Restoration of a Tritium-Aged LaNi4.25Al0.75 Sample

Hydride beds containing LaNi4.25Al0.75 (LANA.75) are used to store significant quantities of tritium. These hydride beds have a limited service life due to radiolytic decay of tritium to 3He within the metal matrix. The crystal structure of the hydride is altered by trapped 3He, which has a very low solubility in the metal. The altered structure induces the formation of a heel of trapped hydrogen isotopes and diminishes the reversible capacity of the hydride. With sufficient tritium exposure, the bed loses the ability to deliver 3He-free tritium, and replacement is needed. Demonstration of a means to regenerate tritium-aged LANA.75 in situ would delay or even eliminate the need to replace lanthanum nickel aluminum (LANA) hydride beds. This paper presents test results obtained during regeneration testing. The efficacy of regeneration testing was evaluated by comparing tritium desorption isotherms collected on the hydride before and after exposure to regeneration conditions. Testing was performed on a bench-scale tritium-aged LANA.75 sample that was previously isotopically exchanged (from tritium to deuterium), passivated, and recovered. Once transferred to a high-temperature test cell, the deuterium heel of the sample was isotopically exchanged with tritium, and a baseline desorption isotherm was collected for comparison purposes. The sample was then heated under vacuum, and comparative isotherms were gathered between regeneration evolutions. Shifts in isotherms show progressive improvements with higher-temperature exposure over the tritium-aged baseline. The heel was significantly reduced, and the reversible capacity of the hydride was essentially restored to near virgin values. For all tested conditions, the plateau pressure remained higher than virgin LANA.75.

Adsorption and Desorption of Tritium in Nuclear Graphite at 700°C: A Gas Chromatographic Study Using Hydrogen

Because of its high content in irradiated nuclear graphite, tritium is treated as one of the most important radionuclides, and it should be carefully decontaminated before the final disposal of nuclear graphite. Tritium has similar chemical and physical characteristics to those of hydrogen; therefore, in this research, the adsorption and desorption of tritium in nuclear graphite using hydrogen were studied. Three kinds of nuclear graphite, IG-110, NBG-18, and NG-CT-10, were used to conduct adsorption and desorption experiments using a new method based on gas chromatography; subsequently, a first-principles calculation on graphene was carried out to simulate the desorption of hydrogen from graphite. The results showed that tritium can be weakly and strongly adsorbed in nuclear graphite. The differences found in the amount of weak adsorption within nuclear graphite were mainly due to the graphite’s porosity and Brunauer-Emmett-Teller surface area, as reported previously in similar research. The mechanism for the strong adsorption was not explained clearly; it could be attributed to the results of a combination of the various physical properties of the graphite, especially the average pore size. The amount of weakly adsorbed hydrogen ranged from 48.4% to 95.2% of the total amount of adsorption for the nuclear graphite working at a temperature of 350°C. The weakly adsorbed tritium easily escaped from the nuclear graphite, indicating that this fraction of tritium would be the main source of pollution during the dismantling or the transportation of decommissioned graphite materials. In addition, the strongly adsorbed hydrogen began to be desorbed when the nuclear graphite was heated over 600°C, and 14% to 71% of the stably adsorbed hydrogen was desorbed when the temperature reached 700°C. A first-principles calculation indicated the activation energy for desorption of tritium from graphene to be about 2.17 eV.

Tritium retention in W plasma-facing materials: Impact of the material structure and helium irradiation

Plasma-facing materials for next generation fusion devices, like ITER and DEMO, will be submitted to intense fluxes of light elements, notably He and H isotopes (HI). Our study focuses on tritium (T) retention on a wide range of samples: first, different types of materials were investigated to distinguish the impact of the pristine original structure on the retention, from W-coated samples to ITER-grade pure samples submitted to various annealing and manufacturing procedures, along with monocrystalline for reference. Then, He and He-D irradiated samples were studied to investigate the impact on He-damages such nano-bubbles (exposures in LHD or PSI-2) on T retention. We exposed all the samples to tritium gas-loading using a gentle technique preventing any introduction of new damage in the material. Tritium desorption is measured by Liquid Scintillation counting (LSC) at ambient and high temperatures (800 °C). The remaining T inventory is then measured by sample full dissolution and LSC. Results on T inventory on He exposed samples highlighted that in all cases, tritium desorption a gas (HT) increases significantly due to the formation of He damages. Up to 1.8 times more T can be trapped in the material through a competition of various mechanisms, but the major part of the inventory desorbs at room temperature, and so will most likely not take part to the long-term trapped inventory for safety and operational perspectives. Unfortunately, investigation of as industrial (used for the making of plasma-facing materials) highlighted a strong impact of the pre existing defects on T retention: up to 2.5 times more T is trapped in as received W compared to annealed and polish W, and desorbs only at 800 °C, meaning ideal material studies may underestimate T inventory for tokamak relevant conditions.

Applicability evaluation of a natural circulation loop to a uranium hydride bed for tritium accountability

A PVT-c method, which is widely used to measure the amount of tritium, is not practical for a UHx bed because the process is time-consuming and includes tritium desorption and transfer processes, which can increase the potential risk of tritium release. In this study, a natural circulation (NC) loop, which can compensate for the issues of a PVT-c method, was numerically designed and evaluated for tritium accountability (TA) of a UHx bed. The numerical evaluation shows that the NC loop can be applied to a UHx bed for TA. Moreover, the characteristics of various parameters including the geometry, and initial pressure of the NC loop, were obtained. These characteristics can be commonly applied to any type of NC loops using heat sources and sinks.