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At present, the biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramic pebbles with high lithium content and crushing load have been viewed as promising tritium breeders for maintaining the safety-state operation of the D-T fusion reactor. The tritium release behavior of Li 2 TiO 3 –Li 4 SiO 4 ceramic pebbles fabricated by centrifugal granulation method was investigated by using the Thermal Desorption Spectroscopy apparatus at Shizuoka University. It can be found that the Li 2 TiO 3 –Li 4 SiO 4 ceramic pebbles show three tritium release peaks, and most of the tritium was released as HTO observed from T-TDS spectra, and the peaks at 581 K and 734 K are mainly due to tritium release from Li 2 TiO 3 in core and shell. The peak at 813 K is mainly deemed as tritium release derived from the Li 4 SiO 4 shell. The content of tritium release from biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramic is up to 11.5 MBq g−1, but the amount of gas species (HT) is only 0.126 MBq g−1. Besides, the isothermal heating experiment showed that the tritium release behavior was subject to diffusion process and de-trapping in defect sites. At the same time, the irradiation defects of E′-center, O-related center, and Ti3+ for breeding material can be found by using an ESR device, and the isochronous heating experiments have manifested that the amount of irradiated defect decreases as heating temperature increases, which indicates that the tritium release behavior is closely related to defect annihilation process. Therefore, the advanced Li 2 TiO 3 –Li 4 SiO 4 ceramic breeder can be fabricated based on tritium release result, which will be conducive to meet the practical application of breeder in the future. [ABSTRACT FROM AUTHOR]

Biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramics have been proposed as advanced tritium breeder for solid-type breeding blankets of fusion reactors. However, the mechanism of tritium release from biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramics remains to be elucidated. In this study, an integrated numerical model for tritium release from biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramics was constructed for the first time. In the proposed model, besides the mass transfer steps for tritium release from single-phase ceramics (e.g. Li 2 TiO 3 and Li 4 SiO 4), a mass transfer step between the interfacial layers of Li 2 TiO 3 and Li 4 SiO 4 was also included to take into account the effect of two-phase interface in Li 2 TiO 3 –Li 4 SiO 4 ceramics. By using this model, numerical simulations of tritium release from Li 2 TiO 3 –Li 4 SiO 4 ceramics with different phase ratio and grain size were performed. The simulation results showed that the mass transfer of tritium atoms between the interfacial layers of Li 2 TiO 3 and Li 4 SiO 4 can effectively enhance tritium release from Li 4 SiO 4 in biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramics. Furthermore, increasing phase ratio of Li 2 TiO 3 and decreasing grain size of Li 2 TiO 3 were both beneficial to tritium release from biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramics at lower temperature. The modeling work consolidated the presumption that the two-phase interface could play an important role in the tritium release process of biphasic Li 2 TiO 3 –Li 4 SiO 4 ceramics. [ABSTRACT FROM AUTHOR]

Understanding the tritium breeding and release behavior of ceramic breeder is crucial for the parameter design of solid blanket in deuterium-tritium (D-T) fusion reactors. Lithium titanate (Li 2 TiO 3) is commonly considered an attractive candidate tritium breeder material. Defect types and densities formed by neutron irradiation at different energies result in diverse tritium release behaviors in Li 2 TiO 3. To understand the mechanism of high-energy neutrons on the tritium release of solid-state breeders, it is crucial to elucidate the tritium production and release behavior of Li 2 TiO 3 under 14 MeV fusion neutron irradiation. In this study, the high-intensity D-T fusion neutron source (HINEG-CAS) was employed to conduct tritium production experiments on Li 2 TiO 3 ceramic breeder samples. Tritium release experiment was also performed on irradiated Li 2 TiO 3 samples using the release system to obtain the temperature-dependent pattern of tritium release induced by fusion neutrons. The results indicated that a limited but visible amount of tritium was released at room temperature. Defect self-healing behavior at RT was seemly observed. The tritium release peak occurred at ∼673 K as the temperature increased. After continuous purging for 4 h at 1073 K, no residual tritium was observed in the irradiated sample, indicating successful collection through bubblers. The total radioactivity of released tritium amounted to 1866.4 Bq, predominantly in the form of tritiated water (HTO: 79.3%). Furthermore, the preliminary analysis of the mechanism behind tritium release has been provided as well. The investigation is conducive to the design optimization of the breeding blanket. • The tritium production and release experiments of Li 2 TiO 3 were conducted using 14 MeV fusion neutron irradiation. • A limited but clearly observable amount of tritium released at room temperature due to the defect self-healing behavior. • The peak of tritium release occurred at approximately 673 K and the primary form was identified as HTO (79.3%). • The influence of water vapor, sensitive tritium measurement method, and heating rate were essential for tritium release. [ABSTRACT FROM AUTHOR]

This paper examines in situ spectroscopic measurements of nuclear-excited plasma of noble gases excited by 6Li(n,α)3H nuclear reaction products in the core of a nuclear reactor. A thin layer of lithium applied on the walls of the experimental device, stabilized in the matrix of the capillary-porous structure, serves as a source of gas excitation. During in-pile tests conducted at the IGR research reactor, thermal neutrons interact via the 6Li(n,α)3H reaction, and the emergent α-particles with a kinetic energy of 2.05 MeV and tritium ions with a kinetic energy of 2.73 MeV excite the noble gas (Ar) medium. The intensity of tritium release from the lithium layer in noble gases was estimated by the intensity of the α-line of the Balmer series of the tritium atom 3Hα (656.2 nm). A tritium release was observed at 710 K due to the beginning of desorption of thermalized tritium atoms dissolved in the liquid phase of lithium. The results are of interest in terms of clarifying the mechanisms and developing models that allow for describing the processes of generation, diffusion, and release of tritium from lithium during neutron irradiation. [ABSTRACT FROM AUTHOR]

We report an experimental study on the tritiding capabilities over a long period of the intermetallic compound Zr(Fe0.5V0.5)2. The study was carried out with the prospect of using the alloy as a chemical converter to reduce HTO. Two identical getter beds, containing 1 gram of alloy in powder form each, were used in the experiments. While one of them was exploited to determine the tritium isotherms of the virgin alloy, the other bed was loaded at 75% of stoichiometry with 354 Ci of tritium and left to age for 1500 days. The bed was then unloaded and the isotherms of the aged alloy were determined twice to check the repeatability. The main results of the work are that, while enthalpy and entropy changes for tritium dissolution at infinite dilution are practically the same for the fresh alloy and the aged alloy, they vary significantly when the isotherms are determined on the aged alloy at a large enough distance of time (one week). This behavior is ascribed to the He3 present in the interstitial sites. However, the fact that the solubility of the alloy decreases with aging suggests that the He3 present either in the interstitial sites or in bubbles subtracts sites for dissolution. Also to be stressed is that in the tritide-forming region, these thermodynamic values decrease with aging in a monotonic way. This different behavior is tentatively explained by invoking the nature of the tritium bond. [ABSTRACT FROM AUTHOR]

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%.

Two-phase ceramics Li2TiO3-Li4SiO4 are one of the potentially promising materials for creating a ceramic blanket for DEMO reactor. However, until now, a limited number of studies have been carried out on the release of tritium from this material under the influence of neutron irradiation, which poses fundamental problems for its application.In this work, the pebbles of two-phase lithium ceramics consisting of 25 mol% Li2TiO3 and 75 mol% Li4SiO4 were studied using thermal desorption spectroscopy (TDS) after their irradiation with neutrons at the WWR-K research reactor. Under the conditions of the irradiation experiment, which lasted 21.5 days at a reactor power of 6 MW, the samples were exposed to thermal neutrons with a flux density of 2·1013n/(cm2 × s). The thermal neutron fluence accumulated as a result of irradiation was 3.7·1019 cm−2.It has been established that tritium comes out mainly in the form of HT molecules and has three visible TDS peaks, which can be described by three rates of the first-order desorption reaction with an activation energy of 90 kJ/mol. Experiments have shown that the main amount of tritium is evenly distributed throughout the pebble’s pores and voids. Its yield is determined by desorption and transfer from inner voids to the boundaries that communicate with the outer surface of the sample. The kinetics of tritium release depends on the number, size and depth of the exit paths of such areas inside the ceramic.

In the current paper, we have studied the impact of microwave (2.45 GHz) plasma treatment of submicronic (250–600 nm) tungsten dust, upon tritium gas retention. Herein, the conducted experiments have emphasized the role of dust treatments in pure hydrogen gas versus hydrogen plasma, before the tritiation process at different pressures. The obtained tritiated dust was analyzed via room temperature desorption and dissolutions. Additionally, Scanning Electron Microscopy and X-ray Photoelectron Spectroscopy analyses were performed to observe the changes induced by the plasma discharge. The results have shown that the Specific Surface Area of dust is enhanced by using microwave hydrogen plasma treatments, resulting in a high tritium gas retention inside the submicronic tungsten dust. [Display omitted] • We studied the 3H retention in W submicronic dust. • One compared 3H retention in untreated/plasma-treated dust. • Microwave plasma dust treatment enhanced 3H retention. [ABSTRACT FROM AUTHOR]

In order to have a deep understanding of the tritium migration and release mechanisms in ceramic breeders, computer simulation of the tritium release behavior of Li4SiO4 ceramic breeder was performed by using an improved tritium release model. The influences of various factors, such as surface adsorbed water, water vapor in the purge gas, hydrogen in the purge gas, grain size and tritium production amount, on the tritium release process were systematically investigated. The simulation results have shown that: 1) The surface adsorbed water and water vapor in the purge gas can remarkably facilitate tritium release as tritiated water (HTO) in low temperature region (typically <450 °C); 2) Adding hydrogen to the purge gas is effective to promote tritium release as tritium gas (HT), but the effectiveness is sensitive to surface adsorbed water and water vapor in the purge gas; 3) Large grain size and tritium production amount can relatively weaken the effects of surface adsorbed water and water vapor in the purge gas. In general, the simulation results were consistent with the observations in tritium release experiments, which were helpful to further elucidate the complex behavior of tritium release from Li4SiO4. [ABSTRACT FROM AUTHOR]

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%. [ABSTRACT FROM AUTHOR]

Biphasic lithium ceramics based on lithium orthosilicate Li4SiO4 and lithium metatitanate Li2TiO3 is one of the most promising materials for breeder blankets of future fusion reactors. One of the important issues of biphasic lithium ceramics application in the fusion reactor blanket is to determine the parameters and mechanisms of tritium transfer within and from the ceramics.This paper continues the analysis of irradiation experiments carried out at the WWR-K reactor (Almaty, Kazakhstan) with a sample of biphasic lithium ceramics Li4SiO4-Li2TiO3 (pebbles of lithium orthosilicate with 35 mol% lithium metatitanate with diameter 250––1250 μm). The section of the experiment in which the reactor was temporarily shutdown for 1.5 h was investigated in detail. During this period of time the sample temperature rapidly decreased from 665 °C to 100 °C, generation of tritium and helium in the lithium ceramic sample ceased, but the desorption of previously generated gases from the ceramic surface continued. The experiments were carried out by the vacuum extraction method.The nature of tritium-containing molecules and helium release for that specified time interval was analyzed. The kinetics of tritium release from ceramics in the experiment during reactor shutdown was simulated and the expression for the effective diffusion coefficient D = 5e-11(m2/s)∙exp(-20(kJ/mole)/RT) was determined. It was suggested that one the most realistic mechanisms for tritium release is the mechanism associated with both diffusion and desorption of tritium from the pebbles surface and release from the open pores of the pebble.This mode of the experiment made it possible to estimate the parameters of tritium release immediately after irradiation, which imitates the conditions of breeding blanket operation in the fusion reactor.

We report an experimental study on the tritiding capabilities over a long period of the intermetallic compound Zr(Fe0.5V0.5)2. The study was carried out with the prospect of using the alloy as a chemical converter to reduce HTO. Two identical getter beds, containing 1 gram of alloy in powder form each, were used in the experiments. While one of them was exploited to determine the tritium isotherms of the virgin alloy, the other bed was loaded at 75% of stoichiometry with 354 Ci of tritium and left to age for 1500 days. The bed was then unloaded and the isotherms of the aged alloy were determined twice to check the repeatability. The main results of the work are that, while enthalpy and entropy changes for tritium dissolution at infinite dilution are practically the same for the fresh alloy and the aged alloy, they vary significantly when the isotherms are determined on the aged alloy at a large enough distance of time (one week). This behavior is ascribed to the He3 present in the interstitial sites. However, the fact that the solubility of the alloy decreases with aging suggests that the He3 present either in the interstitial sites or in bubbles subtracts sites for dissolution. Also to be stressed is that in the tritide-forming region, these thermodynamic values decrease with aging in a monotonic way. This different behavior is tentatively explained by invoking the nature of the tritium bond.

• Most of the adsorbed water on LiAlO 2 was released over around 250 °C. • Most of T was released as tritiated water vapor from the irradiated LiAlO 2 powder. • T release peaks from the irradiated LiAlO 2 were observed at 250, 550, 700, 800 °C. • T was mostly released as gaseous tritium from the preheated LiAlO 2 mixed with zr. • T release peak from the preheated LiAlO 2 mixed with Zr was observed at 600 °C. Tritium production using nuclear reactions of neutrons with lithium in a high temperature gas-cooled reactors has been studied as an external source of fuel tritium in the early stage of fusion reactor operation. In order to control tritium migration throughout the reactor, it is important to understand tritium release behaviors from Zr-containing LiAlO 2 , which are used as tritium producing materials. In this study, tritium release behavior from neutron irradiated LiAlO 2 with and without Zr ware investigated by heating to 900 °C. In the case of heating only LiAlO 2 , most tritium was released in the chemical form of HTO. On the other hand, in the case of heating Zr-containing LiAlO 2 , the chemical form of tritium was mostly HT. This result indicates that even if tritium is released from LiAlO 2 as HTO, it is effectively absorbed by Zr at 900 °C. [ABSTRACT FROM AUTHOR]

During the past decades, fusion reactor fuels such as deuterium and tritium have been extensively investigated due to increasing interest in nuclear fusion energy. Tritium, which is scarce in nature, needs to be fabricated by tritium breeder materials. Among the commonly investigated tritium breeder materials, lithium titanate (Li2TiO3) is recognized as one of the most promising solid tritium breeder materials because of its considerable lithium (Li) atomic density, low activation, excellent chemical stability, and low-temperature tritium release performance. This paper aims to provide a systematic review of the current progress in Li2TiO3 preparation methods as well as the high Li density, tritium release performance, irradiation behavior, and modification technologies of Li2TiO3 pebbles. Li2TiO3 can be synthesized by strategies such as solid-state, sol–gel, hydrothermal, solution combustion synthesis, and co-precipitation methods. Among them, the hydrothermal method is promising due to its simplicity and low cost. Many researchers have begun to focus on composite ceramic pebbles to further improve tritium breeder performance. This will provide a new direction for the future development of Li2TiO3 pebbles. The present review concludes with a summary of the preparation methods currently under development and offers an outlook of future opportunities, which will inspire more in-depth investigation and promote the practical application of Li2TiO3 in this field. [ABSTRACT FROM AUTHOR]

FeCrAl alloys, one of the promising ATF (Accident Tolerant Fuel) claddings, have been widely researched because of their superior oxidation resistance. However, compared with zirconium-based cladding, the higher tritium permeability of FeCrAl alloys could result in a higher tritium concentration in the primary coolant water. Deuterium transport through ferrite FeCrAl alloys compared with 316L-SS was investigated using a permeation apparatus. Deuterium permeability, diffusivity and solubility in FeCrAl alloys were measured. A 2D finite element method model was built based on updated test results to analyze the tritium release from LWRs fuel rods. Deuterium diffusivity in FeCrAl alloys is more than one order of magnitude higher than 316L-SS, but the solubility is much lower. In addition, the impact of all these data on LWRs operation was discussed. The diffusion rate significantly affects the time to reach steady state. Still, even if the diffusion rate is reduced by a factor of 0.01, it does not affect the steady-state tritium permeability from the fuel rods. FeCrAl fuel rods would experience a considerable tritium flux release rate because of the high diffusivity. • Deuterium diffusivity and solubility in FeCrAl alloys were innovatively measured. • A calculation method for tritium permeation from fuel rods was established. • Deuterium diffusivity in FeCrAl alloys is over one order of magnitude higher than 316L-SS. • The tritium diffusion rate has a great effect on the time to reach a steady state. • FeCrAl Fuel rods have a considerable tritium flux release rate due to the high tritium diffusivity in FeCrAl. [ABSTRACT FROM AUTHOR]

This work presents the preliminary experimental data on the study of gas release from two-phase lithium ceramics Li4SiO4-Li2TiO3 under neutron irradiation conditions. Experiments were carried out at the WWR-K research reactor (Almaty, Kazakhstan) for ∼4.3 days. The total neutron fluence during the irradiation was ∼1.8·1019n/cm2. In the course of irradiation, two experiments on ceramics heating during irradiation and two experiments with hydrogen isotopes (H2 and D2) supply into the experimental chamber with the sample were performed at a temperature of 680 °C and reactor power of 6 MW. During the entire irradiation, the gas composition in the continuously evacuated ampoule device with samples was recorded. The obtained dependences of the release of tritium-containing molecules and helium during the experiment were qualitatively analyzed.

In the development of fusion energy, an important task is the study and improvement of tritium production technologies. In this case, one of the most promising materials for tritium generation is lithium ceramics. Considering the importance of the task, numerous studies are aimed at solving the problem of determining the parameters and mechanisms of tritium release in lithium-containing materials. This paper presents the results of a study of tritium release processes from two-phase lithium ceramics of Li4SiO4/Li2TiO3 during reactor irradiation when hydrogen and deuterium are injected into the chamber with irradiated samples. The mechanisms regularities of the tritium yield process in the presence of these isotopes were established. The experiments were carried out in the WWR-K research reactor at a neutron flux density of 5∙1013n/cm2∙s and sample temperatures from 650 to 700 °C.

This paper analyzes part of the reactor experiments on the study of tritium and helium release from promising two-phase lithium ceramic (Li2TiO3 and Li4SiO4) of natural lithium enrichment conducted by vacuum extraction.The basis for such an analysis was a more careful study of the time trend of pressure changes of gases in the chamber with the test samples. In a particular case, it was clearly shown that the pressure fluctuations observed during irradiation for gases with mass number M4 (to which both HT and He molecules correspond) are determined only by He, which leaves the intergranular regions of the ceramic through open channels or cracks. The kinetics of changes in the amount of helium that is released during irradiation was traced and both the rate of helium release and the frequency of emissions were determined. It was assumed that the observed emissions correspond to a certain “formation of free paths” from the internal cavities of the irradiated ceramics into the chamber of the facility. The data obtained for the helium emissions were compared with the release of tritium-containing molecules from the ceramics. The quasi-equilibrium levels of the release of tritium-containing molecules and their dependence on the reactor power were estimated. The release of helium and tritium was compared with the calculated values of the tritium generation rate in the test sample.

Immersion, percolation and tritium release experiments in peat and vermiculite soil samples were performed to analyze their behavior in this widely used medium for plant cultivation. Samples were immersed in tritiated water for 696 h and the isotope exchange capacity evaluated. A vertical flow regime was also considered with analysis for hydraulic conductivity to understand tritium mobility and therefore its availability. Peat soil showed a high tritium retention after percolation, but vermiculite seem to suppress its retention ability. The high moisture and organic content of peat enhanced its isotope exchange capacity. The falling head method was used to numerically evaluate the saturated hydraulic conductivity and outflow flux. Calculated isotope exchange capacity was 4.95×10-2 mol-T 2 O/g for peat and 3.38×10-2 mol-T 2 O/g for vermiculite. The tritium release experiment showed significant release of tritiated carbons in peat. • Tritium is incorporated in peat and vermiculite through an isotope exchange reaction. • HTO (g) released from peat at high temperatures due to tritium in mineral components. • Isotope exchange reaction rate in vermiculite is slower than that in peat. • Soil with water molecules structurally bonded undergo significant tritium exchange. [ABSTRACT FROM AUTHOR]

Lithium-based tritium breeding materials will be adopted to produce tritium in the D-T fusion reactor. Li2TiO3 and Li4SiO4 pebbles have been proposed as candidates in water-cooled ceramic breeder blanket (WCCB) and helium-cooled ceramic breeder blanket (HCCB) respectively. Biphasic ceramics of core–shell Li2TiO3-Li4SiO4 have been considered as advanced breeding materials due to combining the superiority of Li2TiO3 and Li4SiO4. The defects will be introduced into tritium breeders in the operation of fusion reactor. They have important effects on tritium release. It is necessary to carry out the experiment on irradiation defects evolution. The annihilation kinetics of defects induced by γ-ray irradiation were investigated. Fluka and Flair were used to calculate the DPA (displacement per atom) of these irradiated pebbles. EPR (Electron Paramagnetic Resonance) characterization and EasySpin simulation were adopted to analyze the evolution and annihilation kinetics processes of irradiation defects. The concentration of defects decreased as the annealing temperature increased. There were still a certain amount of defects in Li4SiO4 and Li2TiO3-Li4SiO4 when the defects in Li2TiO3 disappeared by annealing. Li2TiO3 pebbles have a better irradiation stability than that of Li4SiO4 pebbles. According to the results of EasySpin simulation, the defect concentration of E’-center and O-related was obtained respectively. The kinetics parameters for the defects of E’-center and O-related center in Li2TiO3 and Li4SiO4 were acquired. The evolution behavior of Ti3+ in core–shell Li2TiO3-Li4SiO4 contributes to the recovery of defects. The correlation between annihilation of irradiation defects and tritium release was discussed. The anti-irradiation damage stability of three kinds of ceramic breeders were evaluated. This work carries out the research on the defect kinetics of the new core–shell structure tritium breeder. Meanwhile, a comprehensive comparison of the annihilation properties of three promising breeders has been made. Li2TiO3 pebbles present excellent tritium release performance due to higher annihilation rate constant. Compared the performances of these breeder pebbles, Li2TiO3 plays a positive role in irradiation tolerance and tritium release performance, and Li4SiO4 has higher lithium density which benefit for tritium production.

The processes of tritium generation and release in lithium-ceramic breeder blankets are affected by many factors, for example, the presence of impurities in the purge gas. As was shown earlier, oxygen, water vapor and hydrogen have the greatest influence on gas release from lithium ceramics. Hydrogen is planned to be specifically added to the purge gas mixture in breeder blanket to facilitate tritium release. The paper presents data of reactor experiments on irradiation of two-phase lithium ceramics (65 mol% Li 4 SiO 4 /35 mol% Li 2 TiO 3) under conditions of supplying various gases and water vapor into the chamber with samples. We analyzed the sections of experiments in which sequential, step-by-step injection of water, oxygen, hydrogen and deuterium vapors into a continuously pumped chamber with samples at steady-state quasi-equilibrium release of tritium from ceramics was carried out. Using a mass spectrometer, the gas composition in the chamber and, in particular, the release features of tritium-containing molecules, were recorded. Based on the obtained data, the mechanisms of tritium interaction with the investigated gases and water vapor on the ceramic surface were determined, as well as their influence on the processes of tritium release from ceramics. [ABSTRACT FROM AUTHOR]

This paper investigates the effects caused by Tritium released from spent nuclear fuel (SNF) stored into a spent nuclear pool made of (reinforced) concrete. The release of tritium from fuel elements to coolant during normal LWRs operation is of 1 % while during 6–20 years of wet storage is assessed to 13–45 %. Permanent Tritium concentration in coolant is assessed to n·109 Bq·dm−3. The water contamination first and the filtration of the concrete pool walls then pose important safety problems to face from both radiological and structural point of view: during Tritium filtration, radiation affects significantly the structural and mechanical properties of concrete due to the breaking of atomic bonds. Under the effect of radiation concrete microstructure modifies; collision determines atom dislocation and so a lattice defect. Ionized rays may cause the decay of free or bonded water leading to the formation of H2 and O2 and produce explosive mixture. The damage from "soft" β-decay of Tritium inside the concrete could be harder due to short track of the particle and the resulting "volume" effect. Neither this effect nor to what extent Tritium movement in/through concrete material would take place over long-term are yet completely studied. Due to Helium (He) formation, when Tritium decays, the restructuring of water molecules with generation of reactive free radicals occurs. We have assessed these effects: the irradiation and accompanying volume expansion, formation of free radicals and corresponded chemical corrosion are not considerably influencing on the structure and sustainability of concrete biological shield. The most significant effect is the radiation breaking of van der Waals bonds that within 10 years can lead to the destruction of about 2 % of concrete. Concrete shield does not provide a complete barrier against Tritium release into the environment. The main threat of Tritium diffusion to coolant in SNF storage pools is caused by filtration of tritiated water through the concrete walls that may provide significant health hazards for employees of the nuclear industry. Results show that β-decay of tritiated water adsorbed in 1 m3 of concrete may destroy about of 0.05 mol (11.4 g) of cement stone within 10 years, which is less than 0.001%. Moreover, with 3. With reference to the 5th power unit of the Zaporizhzhia NPP, the volume of Tritiated water released from the pool is 0.67 m3 annually. show that the tritium waste management concept needs to be revised. • Analytical evaluation of the Tritium released during SNF wet storage. • GIS methodology to quantify the tritium release and environment impact. • Calculation of HTO of the 5th power unit of the Zaporozhye NPP. [ABSTRACT FROM AUTHOR]

Highlights • A two-dimensional tritium transport analysis is performed for water cooled ceramic breeder blanket for CFETR. • The influence of temperature and velocity field on the tritium transport is considered. • The sensitivity analysis about the velocity of purge gas is conducted. Abstract As for the fusion reactor, tritium has received widely attention due to its self-sufficiency and its influence on the safe operation of the fusion reactor. As a result, quantitative analysis of tritium diffusion, inventory, permeation in the blanket is crucial to the design of tritium fuel circulating system and safety issue. In this study, a two-dimensional model for tritium transport in water cooled ceramic breeder (WCCB) blanket is created, and finite element software COMSOL is used to simulate the tritium transport process. The amount of tritium inventory, the tritium permeation into the coolant, the tritium taken by purge gas and the tritium release are calculated. Besides, the concentration profile of tritium and temperature profile in the blanket are given as well as the sensitivity analysis about the velocity of purge gas is conducted which can offer a reference for the design of tritium extraction system. [ABSTRACT FROM AUTHOR]

Abstract The tritium release behavior of the Li 4 SiO 4 pebbles and Li 4 SiO 4 + 5 mol% TiO 2 pebbles with small grain sizes was investigated. The tritium release results of Li 4 SiO 4 pebbles with different grain sizes (0.3 μm and 1.5 μm) indicated that the grain size had little effect on the tritium release regardless of the difference in gas composition of purge gas. Moreover, the tritium release of small grained pebbles was dominated by the desorption process, since the addition of H 2 to purge gas substantially affected the release behavior. The modified Li 4 SiO 4 pebbles with addition of TiO 2 exhibited enhanced water formation capacity due to the increased concentration of active point as the oxygen supplier. Obvious tritiated water release peaks around 690 °C could be observed from the Li 4 SiO 4 + 5 mol% TiO 2 pebbles under 0.1%H 2 +He purge gas. The modified Li 4 SiO 4 pebbles also showed improved tritium release behavior, the tritium release peaks shifted to lower temperatures compared to Li 4 SiO 4 pebbles. [ABSTRACT FROM AUTHOR]

Highlights • A homogenization treatment is necessary to form a single-phase structure in titanium beryllide. • Homogenization causes an enhanced pore formation in titanium beryllide. • Homogenization facilitates tritium release from titanium beryllide. Abstract In the present paper, a homogenization heat treatment at 1473 K for 8 h was carried out with the aim to obtain the single-phase Be12Ti structure in the pebbles of three Be-Ti compositions: Be-7.0Ti, Be-7.3Ti, Be-7.7Ti at.%. After the heat treatment the average porosity in the pebbles increased from 5 to 8% to 36–47%, which resulted in lower amount of tritium/hydrogen retention during the loading. Temperature-programmed desorption tests showed that the performed heat treatment facilitates tritium/hydrogen release from Be-Ti pebbles. Assessment of the effective activation energies of tritium desorption reveals that the tritium release from titanium beryllide occurs much more easier than from pure beryllium. [ABSTRACT FROM AUTHOR]

Abstract This paper presents the results of investigation of three batches of beryllium pebbles with average pebble size of 1.2–1.3 mm and different average grain sizes (13–14 μm, ∼50 μm and ∼615 μm). Microstructure and chemical composition of produced beryllium pebbles are presented as well as packing density and pebble size distribution. The influence of grain size on tritium release and retention in Be pebbles during temperature programmed desorption (TPD) after high-temperature loading of tritium/hydrogen gas mixture are also described. [ABSTRACT FROM AUTHOR]

Li 2 TiO 3 -Li 4 SiO 4 biphasic ceramic pebbles have been developed as an advanced tritium breeder due to the potential to combine the advantages of both Li 2 TiO 3 and Li 4 SiO 4 . Wet method was developed for the pebble fabrication and Li 2 TiO 3 -Li 4 SiO 4 biphasic ceramic pebbles were successfully prepared by wet method using the powders synthesized by hydrothermal method. The tritium release properties of the Li 2 TiO 3 -Li 4 SiO 4 biphasic ceramic pebbles were evaluated. The biphasic pebbles exhibited good tritium release property at low temperatures and the tritium release temperature was around 470 °C. Because of the isotope exchange reaction between H 2 and tritium, the addition of 0.1%H 2 to purge gas He could significantly enhance the tritium gas release and the fraction of molecular form of tritium increased from 28% to 55%. The results indicate that the Li 2 TiO 3 -Li 4 SiO 4 biphasic ceramic pebbles fabricated by wet method exhibit good tritium release property and hold promising potential as advanced breeder pebbles. [ABSTRACT FROM AUTHOR]

The design of future fusion reactors involves the production of tritium inside the breeder blanket. The most promising material for solid breeder blankets is a two-phase lithium ceramic containing orthosilicate Li 4 SiO 4 (LOS) and metatitanate Li 2 TiO 3 (LMT) of lithium in various proportions. Tritium is formed in lithium under neutron irradiation by the reaction 6Li(n,α)T. Further, this tritium is extracted from the blanket with a purge gas and returned to the fusion zone, realizing the concept of a closed fusion cycle. Irradiation under fission reactor conditions is still one of the few available methods for estimating the parameters of tritium generation and release from lithium-containing materials in the "in-situ" mode. This paper presents the results of experiments on neutron irradiation of two-phase lithium ceramics of various ratios (LOS + 35 mol. % LMT (pebble size 250–1250 μm), LOS + 35 mol. % LMT (pebble size 500–710 μm) and LOS + 25 mol. % LMT (pebble size 500–710 μm) at the WWR-K research reactor. Irradiation of each batch of samples lasted from 5 to 22 days. The experiments were carried out by the vacuum extraction method. This paper describes the main methodological aspects of the studies, namely the technical features of four irradiation campaigns, the sequence and scope of the studies. A comparison is also made of the initial sections of reactor experiments for all campaigns, where the reactor was sequentially brought to power, according to which the parameters of the Arrhenius dependence of the effective tritium diffusion coefficients were estimated. [ABSTRACT FROM AUTHOR]

• A 2D numerical simulation model for tritium release from Al 2 O 3 -coated FeCrAl fuel rods was developed. • Sievert's law and the Soret effect equations were coded and studied in the model. • The low release rate indicates that the Al 2 O 3 coating prevents tritium release effectively. • Tritium concentration at the gap is much higher than the inner cladding for Al 2 O 3 -coated fuel rods. • Locally uncoated defects lead tritium flux to bypass Al 2 O 3 coating and be released outside the fuel rod. FeCrAl is one of the ATF cladding solutions due to its excellent oxidation resistance. In this work, a 2D numerical simulation method was developed to study tritium release from the nuclear fuel pellets to the fuel-cladding gap, tritium adsorption on the inner surface of the cladding, and permeation through Al 2 O 3 -coated FeCrAl cladding into the coolant. For Al 2 O 3 -coated fuel rods, based on Sievert's law, tritium concentration at the pellet-cladding gap is much higher than that at the inner cladding surface for most of the time under normal conditions, which slows down the tritium release rate from the fuel rods. With Al 2 O 3 coating, tritium release from the fuel rods remains rising throughout the lifetime. The release rate is calculated to be low at the end of the lifetime, indicating that the Al 2 O 3 coating prevents tritium release. As the coating thickness decreases, the flux of tritium release is found to be increased. Locally uncoated drawbacks lead tritium in the gap to be preferentially adsorbed by the inner surface of FeCrAl, then diffused in FeCrAl, and finally released outside the fuel rod. Still, more actual irradiation operations and test data are needed. [ABSTRACT FROM AUTHOR]

Tritium release behavior from the Li 4 SiO 4 pebbles with high densities (∼96%TD) and large grain sizes (100–300 μm) fabricated by a melt-based method (the M-OSi sample) was investigated through out-of-pile experiments. Another batch of Li 4 SiO 4 pebbles with relatively low densities (∼86%TD) and small grain sizes (10–50 μm) fabricated by a wet method (the W-OSi sample) was used for comparative study. Comparing with the W-OSi sample, the temperature of tritium release from the M-OSi sample was found much higher. Moreover, the fraction of tritium gas released from the M-OSi sample was much larger, especially under helium purge gas. The big differences between the characteristics of tritium release from the two batches of samples can be explained reasonably by the effect of grain size, implying that the grain size played an important role in the tritium release behavior. This study can provide a guideline for optimizing the fabrication process of Li 4 SiO 4 pebbles. [ABSTRACT FROM AUTHOR]

The influence of zirconium (Zr) dopants on tritium release from Li 2 TiO 3 (001) surface was researched by density functional theory (DFT) calculations. The energy barriers for T 2 and T 2 O molecules formation on undoped and doped surfaces were calculated and compared by using climbing-image nudged elastic band (CI-NEB) method. It was found that T 2 O is more easily generated than T 2 on the surfaces. And Zr dopants can significantly promote the production and release for T 2 O on the surface of Li 2 TiO 3. Especially when Zr atoms doping ratio is to be 3.12 atom % (Zr 3), the barrier of T 2 O formation changed from 0.97 eV on undoped surface to spontaneous formation on doped surface. The energy barrier for T 2 O desorption on the surface also decreases to a smaller energy barrier of about 0.42 eV. Although the catalytic effect of Zr doping on T 2 formation on the surface is relatively weak. The electronic characteristics show that the doping of Zr atoms can influence the charge transfer of surface atoms, making OT group more likely to appeal to adjacent T atom, which may be the reason for promoting the formation of T 2 O by Zr dopants. This study can provide useful details for the effect mechanism and catalytic behavior of Zr dopants on the release of tritium from Li 2 TiO 3 surface. Li 2 TiO 3 surface doped with Zr. (a), (c), and (e) Zr i(i = 1, 2, 3) -Li 2 TiO 3 (001) surface. (b), (d), and (f) Initial structures of T-Zr i(i = 1, 2, 3) -Li 2 TiO 3 (001) surface. (b-1), (d-1), and (f-1) Final structures of T-Zr i(i = 1, 2, 3) -Li 2 TiO 3 (001) surface. Blue, yellow, red, green, and purple atoms indicate Li, Ti, O, Zr, and tritium (T) atoms, respectively. [Display omitted] • Catalysis of Zr dopants on production of T 2 O is more significant than that of T 2. • Zr dopants enable the spontaneous formation of T 2 O without overcoming barrier. • Desorption barrier of T 2 O on Zr-doped surface decreases to a smaller energy barrier of about 0.42 eV. • Charge transfer and electron hybridization occur around dopants during the production of T 2 O molecule. [ABSTRACT FROM AUTHOR]

Comparison of different beryllium samples on tritium release and retention properties after high-temperature loading by tritium/hydrogen gas mixture and following temperature-programmed desorption (TPD) tests has been performed. The I-220-H grade produced by hot isostatic pressing (HIP) having the smallest grain size, the pebbles of irregular shape with the smallest grain size (10–30 μm) produced by the crushing method (CM), and the pebbles with 1 mm diameter produced by the fluoride reduction method (FRM) having a highly developed inherent porosity show the highest release rate. Grain size and porosity are considered as key structural parameters for comparison and ranking of different beryllium materials on tritium release and retention properties. [ABSTRACT FROM AUTHOR]

The behavior of tritium release from Li 4 SiO 4 ceramic pebbles in high magnetic field (MF) was investigated by temperature programmed desorption (TPD). Two batches of Li 4 SiO 4 pebbles produced by wet method were used as the experimental samples, one batch with an average pebble diameter of 0.8 mm (the SMALL samples), and the other 1.2 mm (the BIG samples). A superconducting magnet was applied to generate MF up to 7 T in the sample area during annealing. For both batches of samples, the tritium release curves within and without MF showed very similar characteristics, indicating that the effect of high MF on tritium release behavior is not significant. The tritium release peaks for the BIG samples were observed at much lower temperatures than that for the SMALL samples, even though the grain sizes of the BIG samples are much bigger than that of the SMALL samples. It is considered that surface desorption process dominates the overall tritium release behavior in this work, which probably weakened the MF effect. [ABSTRACT FROM AUTHOR]

The correlation between water desorption and tritium release from Li 4 SiO 4 pebbles was studied by temperature programmed desorption. The released water and tritium from irradiated samples were monitored simultaneously. The main peak for tritium release from the irradiated samples that were exposed to air for more than a month, was shifted from 500 to about 250 °C, as compared to that from the unexposed samples. The peak temperatures for water desorption and tritium release overlapped very well, suggesting a strong correlation between the two processes. Accordingly, a two-step mechanism, involving isotope exchange between the tritium trapped on the grain surface and the surface hydroxyls (–OH), and subsequent desorption of tritiated water through recombination of the –OH/–OT groups, was proposed to explain the tritium release behavior for the air-exposed samples. It is believed that the formation and desorption of surface hydroxyl groups at 200–300 °C can affect the behavior of tritium release from Li 4 SiO 4 significantly. [ABSTRACT FROM AUTHOR]

The temperature of Li 2 TiO 3 pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li 2 TiO 3 pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li 2 TiO 3 pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li 2 TiO 3 layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H 2 /He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H 2 O/He gas (H 2 O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H 2 , it is indicated that the increasing tendency of HT release is similar to that of the dry H 2 /He. [ABSTRACT FROM AUTHOR]