Tritium Release Behavior Research Articles

Tritium release performance of biphasic Li2TiO3–Li4SiO4 ceramic pebbles fabricated by centrifugal granulation method

At present, the biphasic Li2TiO3–Li4SiO4 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 Li2TiO3–Li4SiO4 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 Li2TiO3–Li4SiO4 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 Li2TiO3 in core and shell. The peak at 813 K is mainly deemed as tritium release derived from the Li4SiO4 shell. The content of tritium release from biphasic Li2TiO3–Li4SiO4 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 Li2TiO3–Li4SiO4 ceramic breeder can be fabricated based on tritium release result, which will be conducive to meet the practical application of breeder in the future.

Development of a tritium release system for ceramic tritium breeder pebbles irradiated by high intensity DT fusion neutron Generator at INEST

For ceramic breeder pebbles used in fusion energy, the tritium release data under fusion neutron irradiation is very important and valuable. In this study, a tritium release system for ceramic tritium breeder is introduced, which is matched with the High Intensity DT Fusion Neutron Generator (HINEG-CAS). The HINEG-CAS has been built and operated by the Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences (INEST, HFIPS, CAS), achieving a fusion neutron yield of 1012n/s since 2016. Innovative design and fabrication of sample container enables minimization of the unexpected tritium loss. The dedicated tritium release system can monitor the tritium release behavior over time for both tritiated water and tritium gas, using the bubblers-based tritium trapping system with automatic replacement technology, coupled with highly efficient catalytic oxidants and liquid scintillation counter (LSC). We have performed a tritium release experiment with the Li2TiO3 pebbles as ceramic tritium breeder, and obtained the tritium releasing data at room temperature over time and clarified the tritium chemical form in recovery. It is concluded that this newly established system can serve as a good sharing platform for evaluating prime candidate breeder materials, as well as facilitating detailed design or optimization of tritium involved systems (such as the tritium extraction system).

Modeling of tritium release behavior of biphasic Li2TiO3–Li4SiO4 ceramics

Biphasic Li2TiO3–Li4SiO4 ceramics have been proposed as advanced tritium breeder for solid-type breeding blankets of fusion reactors. However, the mechanism of tritium release from biphasic Li2TiO3–Li4SiO4 ceramics remains to be elucidated. In this study, an integrated numerical model for tritium release from biphasic Li2TiO3–Li4SiO4 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. Li2TiO3 and Li4SiO4), a mass transfer step between the interfacial layers of Li2TiO3 and Li4SiO4 was also included to take into account the effect of two-phase interface in Li2TiO3–Li4SiO4 ceramics. By using this model, numerical simulations of tritium release from Li2TiO3–Li4SiO4 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 Li2TiO3 and Li4SiO4 can effectively enhance tritium release from Li4SiO4 in biphasic Li2TiO3–Li4SiO4 ceramics. Furthermore, increasing phase ratio of Li2TiO3 and decreasing grain size of Li2TiO3 were both beneficial to tritium release from biphasic Li2TiO3–Li4SiO4 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 Li2TiO3–Li4SiO4 ceramics.

HICU PIE results of neutron-irradiated lithium metatitanate pebbles

Lithium metatitanate (Li2TiO3) pebbles were irradiated with neutrons within the HICU (High neutron fluence Irradiation of pebble staCks for fUsion) experiment to investigate their material properties and tritium release behaviour in a post-irradiation examination (PIE). The irradiation temperature is the most significant influence on the material. Besides a higher irradiation temperature, a higher initial Li–6 content tends to lead to an increased tritium generation resulting in the formation of the secondary lithium depleted phase Li4Ti5O12. A pre-compaction of the pebble bed does not have an apparent influence on the material properties. Tritium is released as semi-tritiated water and semi-tritiated hydrogen (HTO and HT) and its release is enhanced at elevated irradiation temperatures.

In-pile tritium release behavior and the post-irradiation experiments of Li4SiO4 fabricated by melting process

Understanding the tritium release and retention behavior of candidate tritium breeder materials is crucial for breeder blanket design. Recently, a melt spraying process was developed to prepare Li4SiO4 pebbles, which were subsequently subjected to the in-pile tritium production and extraction platform in China Mianyang Research Reactor (CMRR) to investigate their in-situ tritium release behavior and irradiation performance. The results demonstrate that HT is the main tritium release form, and adding hydrogen to the purge gas reduces tritium retention while increasing the HT percent in the purge gas. Post-irradiation experiments reveal that the irradiated pebbles darken in color and their grains swell, but the mechanical properties remain largely unchanged. It is concluded that the tritium residence time of Li4SiO4 made by melt spraying method at 467 °C is approximately 23.34 h. High-density Li4SiO4 pebbles exhibit tritium release at relatively low temperatures (＜600 °C) that is mainly controlled by bulk diffusion. The diffusion coefficient at 525 °C and 550 °C is 1.19 × 10−11 cm2/s and 5.34 × 10−11 cm2/s, respectively, with corresponding tritium residence times of 21.3 hours and 4.7 hours.

Computer simulation of tritium release behavior of Li4SiO4 ceramic breeder with an improved packed bed model

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.

Tritium release behavior from neutron-irradiated Li₂+xTiO₃+y with 20wt% Li₂ZrO₃ pebbles under different atmospheres

• LTZO had versatile water vapor release mechanisms that contributed to tritium release. • LTZO had an indigenous peak expected to result from the existence of ZrO2. • The surface reaction contributed the tritium release property from LTZO. • No significant tritium left in LTZO was observed after 900 °C heating. Solid tritium breeding materials are essential for the fusion reactors' fuel cycle. Hence, it is necessary to investigate how the bred tritium releases from them carefully. Li 2+x TiO 3+y with 20 wt% Li 2 ZrO 3 (LTZO) ceramic pebbles were recently developed by QST as an advanced tritium breeder. However, the tritium release behavior has not been fully understood yet. Considering the upcoming high-flux irradiation test and downstream tritium separation and purification system design and permeation loss, further investigation on water vapor release and tritium release properties from LTZO by isokinetic heating were carried out in this study. As a result, it was found that LTZO had diverse vapor release mechanisms, including the same ones as ordinary Li 2 TiO 3 ceramic pebbles and the ones derived from TiO 2 and ZrO 2 . These results suggested that HTO release would last until the water formation reaction between TiO 2 and ZrO 2 , and H 2 in the gas phase ceased. Regarding the tritium release property under different purge gas species, it was experimentally clarified that the surface reaction was the rate-determining step like the existing Li 2 TiO 3 pebbles. In the case of H 2 purging, chemically adsorbed water vapor release and thermal decomposition of LiOH were recognized. Residual tritium in the LTZO pebbles was never observed after 900 °C heating.

Tritium release behavior from neutron-irradiated FLiNaK mixed with Ti powder

• FLiNaK mixed with Ti powder was irradiated by mainly thermal neutrons. • Amounts of TF, T2 and T2O released from the FLiNaK by heating were investigated. • Chemical form ratio of released tritium was approximately TF:T2:T2O = 5:70:25. • Approximately 70% of tritium was released as T2O from the FLiNaK exposed to air. • Ti powder mixed to the FLiNaK contributed to the suppression of tritium release. FLiNaK is used as a simulant of FLiBe and FLiNaBe which are promising candidates of liquid blanket material in fusion reactors. In this study, to obtain useful knowledge for tritium release behavior from molten salts such as FLiBe and FLiNaBe, tritium release experiments were carried out for neutron-irradiated FLiNaK containing Ti powder. The solid-state samples were irradiated by mainly thermal neutrons at Kyoto University Research Reactor, and tritium release behavior as TF and T 2 by heating in Ar gas flow was observed and additionally total amount of released T 2 O was obtained. Most of the tritium generated inside the FLiNaK was released to the gas phase even when the FLiNaK was in a solid state at a temperature below its melting point. From the molten FLiNaK unintentionally exposed to air, approximately 70% of tritium was released as tritiated water vapor. The amount of TF released from the molten FLiNaK was considerably less than that of T 2 . This result is consistent with the tendency of tritium release from neutron-irradiated FLiNaBe in our previous study. The ratio of chemical form was approximately TF:T 2 :T 2 O = 5:70:25. It was experimentally confirmed that the Ti powder added to the salt contributes to the suppression of tritium release from the molten salt.

Study on the chemical compatibility between Li2TiO3 ceramic pebbles and diverse advanced structural materials

AbstractThe tritium breeder and structural materials are necessary components in the blanket to realize tritium (T) self‐sustainment of nuclear fusion. The long‐term exposure between tritium breeders and structural materials will cause surface corrosion in irradiation environments and then further affect the tritium release behavior. In this study, chemical compatibility between Li2TiO3 ceramic pebbles and advanced structural materials was studied systematically at 700 °C for 300 h under He+0.1 % H2 environment, respectively. The color of the Li2TiO3 ceramic pebbles changes from white to dark grey and black. Moreover, the grain size of Li2TiO3 ceramic pebbles increases to more than 5 μm, and the crushing load decreased slightly. For the structural materials, the Al‐rich oxide layer with about 188.7 nm of 14Cr‐5Al oxide dispersion strengthened (ODS) steel and Cr‐Fe rich oxide layer with about 1.04 μm of 14Cr‐ODS steel were observed on the cross‐section. The effective diffusion coefficient of O element in Li2TiO3 ceramic moved into ODS steel at 700 °C was calculated to be 3.3 × 10−16cm2/s and 1.02 × 10−14 cm2/s. Unfortunately, when SiC ceramics were contacted with the pebbles, the crystal phase transformed into SiO2, which severely limits its application. Therefore, these results will provide guidance for the selection of structural materials in the future.

Effects of water adsorption on tritium release behavior of Li4TiO4 and Li4TiO4-Li2TiO3 core-shell structure breeding ceramics

Li4TiO4 and Li4TiO4-Li2TiO3 core-shell breeding ceramics were considered as promising tritium breeding materials in fusion reactor blanket, and the effect of adsorbed water on the tritium release behavior of these two types of materials was investigated and compared in this study. All samples were first irradiated by thermal neutrons at Kyoto University, followed by out-of-pile tritium release experiments at Kyushu University. It was found that both Li4TiO4 and Li4TiO4-Li2TiO3 core-shell breeding ceramics have two tritium trapping sites, mainly due to the overlapping of the tritium release temperature of Li2TiO3 with the second tritium release site of Li4TiO4. Besides, it was observed that the adsorbed water can slightly reduce the tritium release temperature of the two types of tritium breeding materials. More importantly, the effect of adsorbed water on tritium release form (HT/HTO ratio) was quantitatively studied. The results demonstrate that water adsorption has a significant effect on the tritium release form of the breeding ceramics, and the core-shell structure can inhibit the hygroscopicity of Li4TiO4 to increase the fraction of tritium release in the HT form.

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.

Densification behavior and properties of Li4SiO4 ceramic breeder with the addition of SiC as a sintering aid

Lithium orthosilicate (Li4SiO4) is regarded as a vital tritium breeder candidate due to its favorable properties, such as high lithium contents, superior tritium release behavior, and irradiation stability. However, the poor crushing load and low thermal conductivity of the Li4SiO4 ceramic pebbles fabricated via existing methods will lead to the blocking of the tritium release channels, giving rise to thermal stress concentration. To overcome the above shortcomings, the SiC was employed as sintering aid to improve Li4SiO4 ceramic pebbles via simple wet method in this paper. The effects of SiC content on the shrinkage, internal structure evolution, and crushing load of the Li4SiO4 ceramic pebbles are comprehensively investigated, and the densification process is analyzed in detail. It was found that the addition of SiC improves the crushing load (54 N) significantly owing to decreasing the densification activation energy. Moreover, the Li4SiO4 ceramic pebbles with excess Li2CO3 further displays excellent properties at 800 °C, such as the relative density of 90.3%, the crushing load of 65.3 N, and the thermal conductivity of 1.97 W/(m·K) respectively, which is much higher than that of traditional methods (at least 1000 °C). The Li4SiO4 ceramic pebbles with homogeneous micro-pore distribution, the porosity of 7.3% were further measured by using the X-ray computer tomography technique. Therefore, this proposed tactic has great prospects for the service performance of tritium breeding materials in the blanket.

Investigation on effects of tritium release behavior in Li4SiO4 pebbles

For deuterium–tritium fusion reactor, the fuel of tritium is not available naturally. Tritium should be produced by the reaction of Li (n, ɑ) T. Li4SiO4 is one of promising candidates due to its high lithium density. In present work, effects of tritium release behavior from Li4SiO4 have been investigated and compared. The main release peak at lower temperature was attributed to tritium located on/near the grain surface and chemical adsorption on pore surface. The small peak at higher temperature was designated as tritium diffusion from grain including lattice and boundary. The tritium release behavior was simulated by diffusion model and surface reaction model. Tritium behavior is controlled by both diffusion and surface reaction for different stages. Based on the desorption theory, the kinetic parameters were obtained. The effects on tritium release behavior were studied. The porosity has significant effects on tritium release. Higher porosity has large specific surface providing more activation sites and fast channels for tritium release which results in lower temperature of tritium release. The main tritium release form was considered as tritium water. The release peak of deuterium moved towards lower temperature after adsorbing water vapor for 1800 h. It was demonstrated that adsorbing water was beneficial for tritium release due to isotope exchange. Compared the effects of porosity and water vapor, higher porosity has larger effects on tritium release in the present work. Based on the work, the tritium release is controlled by both diffusion and surface reaction at different stage and affected by comprehensive effects including porosity and adsorption water.

The influence of the long-term heating under H2 atmosphere on the tritium release behavior from the neutron-irradiated Li2TiO3

Solid tritium breeding materials are expected to be used in high-temperature conditions for a long time in a fusion DEMO reactor. Thus, it is important to understand how bred tritium releases from the long-term heated material under the environment as close to a DEMO condition as possible to establish a tritium fuel cycle and keep it safe. In this work, the tritium release behavior from the irradiated Li2TiO3 pebbles that was preheated for 720 h at most was observed by heating at 1000 °C under 1000 Pa H2/Ar gas flow. The release peaks of HTO and HT were observed around 300 °C due to the desorption of the chemisorbed water. Also, the broad HTO peak was observed in a higher temperature region despite purging H2/Ar gas. This result suggests that this tritium was released without exchanging with H2 but with combining with oxygen in the pebbles. Moreover, the released tritium amount decreased as the pre-heating time. Finally, the amount of tritium that could not be released by the heating experiment was quantified by dissolving samples with an acid solution. Besides, the total tritium release ratio was discussed.

Tritium release behavior from neutron-irradiated FLiNaBe mixed with titanium powder

FLiNaBe is a promising liquid blanket material for a fusion reactor. Since hydrogen solubility for FLiNaBe is low, one concern is the permeation of bred tritium. In order to increase effective solubility of hydrogen, the addition of Ti powder was proposed. In this study, the influence of Ti addition on the behavior of tritium generated inside FLiNaBe by neutron irradiation was investigated. The samples of FLiNaBe and FLiNaBe mixed with Ti of 5 wt% and 0.5 wt% were prepared and the solid-state samples were irradiated by thermal neutrons at Kyoto University Research Reactor. The neutron-irradiated samples were heated to 700 °C in an Ar gas flow. The released tritium was collected by a water bubbler and the temporal variation of TF and HT release was observed. Initially, the most of tritium was released as HT from both FLiNaBe with and without Ti but eventually, the most of released tritium was HT after long time heating from FLiNaBe with Ti. This result suggests that the addition of Ti can decrease the release of TF very much. The amount of tritium released from FLiNaBe with Ti was lower than that from FLiNaBe without Ti. This result suggests that the addition of Ti can suppress the release rate of tritium from FLiNaBe.

Observation of specific optical phonon modes dominating Li ion diffusion in γ-LiAlO2 ceramic

In γ-LiAlO2 ceramic, Li ion diffusion plays a key role both in tritium release as a solid tritium breeder material in nuclear fusion and in phase stability as a matrix material in molten carbonate fuel cells (MCFC). Yet fundamental understanding of the diffusive process in γ-LiAlO2 is still missing, especially considering the interaction of the lattice dynamics and Li ion motion. Herein, we demonstrated that two specific optical phonon modes are coupled with Li ion diffusion in γ-LiAlO2, by investigating the temperature-dependent lattice dynamics via Raman scattering experiments, as well as first-principles calculations and neutron diffraction experiments (ND). The high-temperature ND experiments showed that γ-LiAlO2 undergoes partial Li ion disordering in Li sublattice at high temperature. Notably, the lattice dynamics studies demonstrated that B1 mode at 226 cm−1 and A1 mode at 263 cm−1 are responsible for Li ion motion through the oscillation of Li–O in <010> and <001> direction, respectively. Our results further suggest that the behavior of tritium release and phase stability in γ-LiAlO2 at elevated temperature may be controlled by tuning the two specific phonon modes through photon/electron-phonon coupling.

Tritium release behavior of Li2TiO3 and 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles fabricated by microwave sintering

Li2TiO3-Li4SiO4 biphasic ceramic is considered as an advanced tritium breeder material for its potential to combine the advantages of both Li2TiO3 and Li4SiO4. In this study, Li2TiO3 and 2Li2TiO3-Li4SiO4 biphasic pebbles with promising properties were fabricated by microwave sintering. The 2Li2TiO3-Li4SiO4 biphasic pebbles displayed porous structure with a grain size of 170 nm. Tritium release behavior for the microwave sintered ceramic breeder pebbles was investigated. It was found that the microwave sintered Li2TiO3 pebbles show the same release behavior with that of conventional sintered pebbles due to the similar microstructure and density properties. Besides, the tritium release results indicated that high closed porosity may lead to the increase of release temperature. The 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles hold satisfactory tritium release behavior with a low release temperature around 450 °C. According to the tritium release results, it could be concluded that the interfaces between Li2TiO3 and Li4SiO4 can improve the tritium release behavior by promoting the surface reaction.

Exhaust behavior of tritium from the large helical device in the first deuterium plasma experiment

ABSTRACT The tritium exhaust behavior from the Large Helical Device (LHD) was observed in the first deuterium plasma experimental campaign. Tritium in the exhaust gas was monitored at the conducted by use of the ionization chamber and water bubbler system with the discrimination of chemical forms. The observation results indicated that (i) tritium on the surface of the first wall and divertor tiles as plasma facing components was released by the hydrogen isotope exchange reaction of the glow discharge cleaning operation and the diffusion-limited process was suggested in the tritium release behavior from the bulk, (ii) the amount of tritium release from the LHD vacuum vessel was about one-third of the produced tritium and the mostly produced tritium was still retained at the end of plasma experimental campaign, (iii) the ratio of exhausted tritium from the LHD vacuum vessel was larger than that in the case of JT-60U based on the carbon materials as the plasma-facing components. It indicated that the tritium inventory would be reduced and controlled by the kind of plasma-facing materials.

Comparison of tritium release behavior in Li2TiO3 and promising core-shell Li2TiO3–Li4SiO4 biphasic ceramic pebbles

In a D-T fusion reactor, self-sufficiency of tritium is one of the critical issues to maintain steady-state operation of the reactor. Li2TiO3 and core-shell Li2TiO3–Li4SiO4 biphasic pebbles have been irradiated by thermal neutrons in Kyoto University. T-TDS (Tritium - Thermal Desorption Spectroscopy) was performed in Shizuoka University. There is one main tritium trapping site in neutron irradiated Li2TiO3. However, three tritium trapping sites are introduced in neutron irradiated core-shell Li2TiO3– Li4SiO4. The two peaks at lower temperature agree with that of Li2TiO3 which can be attributed to tritium release from Li2TiO3 including shell and core. The peak at higher temperature is mainly considered as tritium release from Li4SiO4. The amount of tritium released from core-shell Li2TiO3–Li4SiO4 is slight higher than that of Li2TiO3. The kinetic parameters of tritium diffusion in Li2TiO3 have been obtained. In the case of Li2TiO3 and core-shell Li2TiO3–Li4SiO4, almost all the tritium is released as HTO observed from T-TDS. Purge gas has little effect on tritium release, indicating tritium release from Li2TiO3 was mainly controlled by tritium diffusion in Li2TiO3 grain.

Tritium release property of Li2TiO3-Li4SiO4 biphasic ceramics

The tritium release experiments of Li4SiO4 pebbles, Li2TiO3 pebbles and Li4SiO4-Li2TiO3 biphasic pebbles were conducted to investigated the tritium release behavior of Li4SiO4-Li2TiO3 biphasic pebbles. The results indicate that the biphasic ceramic pebbles with uniform structure and small grain size exhibit good tritium release performance. The tritium release temperature is around 450 °C, and most of tritium is released in the form of tritiated water. The tritium release behavior of Li4SiO4-Li2TiO3 biphasic pebbles is controlled by the release of both Li2TiO3 and Li4SiO4, and the interfaces between Li2TiO3 and Li4SiO4 can promote the tritium release. Besides, the surface reaction is the key process affected the tritium release of the Li4SiO4-Li2TiO3 biphasic pebbles, because the release behavior is significantly affected by the composition of purge gas. Under 0.1% H2+He purge gas, with the combined effect of the two-phase interface and the isotope exchange reaction, the tritium release properties of the biphasic pebbles are promoted.