Tritium Breeding Materials Research Articles

Comparative Study on the Reduction of Tritium Breeding Ratio Caused by Inventory Changes of a Solid-State Tritium Breeding Blanket in a Fusion Demonstration Reactor Using MCNP and FISPACT-II

The self-sufficient supply of tritium in the deuterium-tritium nuclear fusion demonstration reactor (DEMO) is a fundamental design requirement. But, it is hindered by depletion of tritium breeding materials resulting in reduction of tritium breeding ratio (TBR) less than the initial design value especially in the solid-state tritium breeding blanket (TBB) of the DEMO. Unlike the liquid tritium breeding blanket of DEMO, compensation measures of the depleted breeding material in the solid-state TBB will be its substitution depending on the reduction rate of TBR. To estimate the replacement period of the solid-state TBB, it is required to estimate the reduction rate of TBR according to the operation conditions of the DEMO and the physical configuration of a solid-state TBB. In this study, the representative simulation codes, MCNP and FISPACT-II, are used for assessment of the reduction rate of TBR with the benchmarking model which is modified from the one poloidal segment of the TBB in the Korean-DEMO. After 3 full power-year operations with the neutron irradiation with the benchmarking model, the TBR simulated by MCNP is reduced to 96.84% of the initially calculated TBR, but the TBR calculated by FISPACT-II is reduced to 90.57% of the initially calculated TBR.

High pressure phase transition and its influence on structural, mechanical properties and elastic anisotropy of [formula omitted]: An ab-initio study

Lithium carbide (Li2C2), a promising tritium breeding material with potential applications in magnetic confinement-based fusion reactor blankets, has drawn significant attention due to its recent utilization in lithium-ion batteries. In the current study, we employed first-principles calculations with van der Waals (vdW) force correction to explore the structural and mechanical properties of Immm and Pnma phases of Li2C2 subjected to external hydrostatic pressure. It is demonstrated that inclusion of vdW effects accurately predicts Immm→Pnma phase transition at 16 GPa, aligning closely with experimental transition pressure. In contrast, PBE-GGA calculations underestimate the same by about 21%. Current paper is the first comprehensive study of pressure evolution of mechanical properties and strength determining parameters of Li2C2. Interestingly, this inherently brittle material exhibits ductile characteristics beyond 7.5 GPa. The paper also offers a quantitative comparison of its strength parameters with other prominent tritium breeding materials. Finally, the strong directional dependence due to the presence of C2 dimer is investigated by analyzing the pressure variation of elastic anisotropy. The study has its relevance in assessing structural integrity of Li2C2 during inadvertent buildup of tritium and helium that would generate stress in the solid breeder.

Fabrication and characteristics of Li2TiO3 pebbles manufactured by using powder injection molding (PIM) process

The solid type of tritium breeder materials is usually used in a pebble form because of the high packing factor, efficiency of tritium extraction, and advantage of thermal conduction between the structural materials. The breeder pebbles are required a good sphericity and uniform microstructure from the viewpoint of a packing factor and a release of bred tritium. Therefore, in this study, the lithium metatitanate (Li2TiO3) pebbles have been fabricated by powder injection molding process in order to secure a high sphericity of pebbles. In addition, the physical, mechanical, and thermal properties of the pebbles have been investigated. The Li2TiO3 green pebbles have been fabricated by using the powder injection molding process (PIM process). The diameter and sphericity of green pebbles were about 1.000 mm and 1.001, respectively. Then, the green pebbles were sintered at 1000 °C for 3 hr in air atmosphere. The diameter of sintered pebbles is about 0.95 mm. The volume shrinkage ratio was about 16 % during the sintering process. The uniform microstructure with open pores on the inside and outside of the sintered pebbles was observed.

Reaction kinetics of Li4SiO4 and Li4Ti5O12 in biphasic breeder ceramics after lithium burn-up

Advanced ceramic breeder (ACB) pebbles consisting of the two phases Li4SiO4 and Li2TiO3 serve as the EU reference tritium breeding material. In the present study, the phase stability with regard to a lithium-burn-up is investigated. The loss of lithium due to its transmutation will lead to a compositional change of the breeder material. Li4SiO4 and Li2TiO3 will partly be transformed to the lithium-poorer phases Li2SiO3 and Li4Ti5O12, respectively. However, from the phase diagram it is anticipated that only Li2SiO3 will occur as a third phase after a lithium burn-up as the generated Li4Ti5O12 is expected to further react with Li4SiO4 to form again Li2TiO3 and more Li2SiO3. Therefore, the reaction between Li4SiO4 and Li4Ti5O12 and its temperature dependence were investigated in detail and its reaction kinetics were determined. It was found that this diffusion-controlled reaction with an activation energy of 75 kJ/mol (0.78 eV) will slow down at low irradiation temperatures and may not take place at very low temperatures.

Molecular dynamics simulations of the effect of porosity on heat transfer in Li2TiO3

Heat transfer is a key consideration in the development of tritium breeder blankets for future fusion reactors. For solid tritium breeder materials there is a fine balance to be struck between high levels of porosity to encourage tritium release and minimising it to maintain the thermal and mechanical properties. Therefore, in this work we employ molecular dynamics simulations to understand how the introduction of porosity influences the thermal conductivity of lithium metatitanate ceramic breeder material. Our simulations predict that increasing the porosity leads to a decrease in the thermal conductivity which is in good agreement with previous experimental observations. By contrast, we do not observe the increase in the thermal conductivity at high temperatures, that is observed in some experiments. We argue that this increase is a consequence of sintering or some other modification of the experimental sample rather than a fundamental change in the heat conduction mechanism in the crystal matrix.

An apparatus to measure thermal conductivity of ceramic pebble beds under uniaxial compressive stress

Pebble beds have various applications in many industrial and research sectors. Lithium ceramic pebble beds have specific application in fusion reactors as one of the most promising candidates for tritium breeding material. During the operation of the fusion reactor, these pebble beds will have to sustain various thermal and mechanical stresses. Therefore, it is necessary to do thermal characterisation of these pebble beds under different stress conditions and generate sufficient data on their thermal properties, which is essential for designing breeder units of fusion reactors. In this work, an apparatus has been designed to evaluate the thermal conductivity of ceramic pebble beds under compressive stress. It can be mounted to any standard universal testing machine with the necessary adjustments. The apparatus is based on the transient hot-wire method, widely used to evaluate the thermal conductivity of porous materials. A platinum wire with a diameter of 0.5 mm was used as a hot-wire. Instead of utilising a thermocouple, the 4-wire approach was used to obtain the temperature. The transient temperature response of the hot-wire was obtained by applying a 7A constant current to it and monitoring the voltage drop along its 40 mm length. The thermal conductivity of the water gel was measured to assess the accuracy of the apparatus; the results showed a less than 1.5 % variation from the reference value. In this study, the thermal conductivity of Li2TiO3 pebbles was measured in an air and helium environment in different stress conditions. In an air environment, measurements were done at atmospheric pressure, whereas in a helium gas environment, they were performed at gauge pressures of 0.05 bar, 0.5 bar, and 1 bar. The measurements were conducted in two scenarios: (i) constant stress condition at 0, 3, and 6 MPa, and (ii) Cyclic loading-unloading of 0.037–3 MPa and 0.037–6 MPa. It was found that the pebble bed's thermal conductivity rose in tandem with increased compressive stress in both the helium and air environments. Pebble bed’s thermal conductivity was found to be higher in a helium gas environment than in air. The thermal conductivity of the pebble bed increased slightly with rising helium gas gauge pressure from 0.05 bar to 1 bar. This work provides a design to incorporate 4-wire temperature measurement in the hot-wire method for thermal conductivity measurement of compressed pebble beds. It also contributes to the thermal conductivity data of Li2TiO3 pebbles in cyclic stress conditions in helium and air environments.

Design and Development of Hydrogen Isotopes Extraction System at IPR

Tritium breeding, recovery and safe storage are very important for achieving self-sustainability in a nuclear fusion reactor. Indeed, it is one of the essential task in terms of safety and fuel cycle aspects. Usually, tritium breeding is performed in the blanket module. In Indian blanket module, there are mainly two types of tritium breeder materials, one is Lithium based ceramic pebbles Li2TiO3 (Lithium Titanate) as solid breeders and second is the lithium based alloy Pb-Li (Lead Lithium) as a liquid breeder. Tritium bred in solid breeders is extracted with purge helium gas, whereas tritium bred in liquid Pb-Li needs suitable gas-liquid contactor. It should be noted that, tritium recovery (i.e.,-extraction) from Pb-Li is very challenging due to low solubility of tritium in liquid Pb-Li and no such separation technology readily available commercially. Therefore, the necessity for developing a reliable and efficient tritium recovery technology (i.e.,- extractor) for liquid Pb-Li is of key interest.In present work, design and development of Hydrogen Isotopes Extraction System (HIES) for hydrogen recovery from liquid Pb-Li is discussed. As hydrogen and tritium are isotopes, hydrogen is used instead of tritium in present system due to several safety concerns with tritium. Recovery of hydrogen from liquid Pb-Li is carried out using structured packing as a Gas-Liquid Contactor (GLC). The design details of gas-liquid contactor (i.e.-extractor column) and various process tanks of HIES are discussed. In present design, configuration of vertically mounted extractor column makes overall HIES relatively tall (∼ 3 m height) and use of liquid Pb-Li as process fluid makes the system heavy weight (∼1500 kg) too. In addition to this, system needs to be operated above Pb-Li melting temperature (> 235 C) to keep Pb-Li in liquid form. Therefore, thermal analysis of piping as well as seismic analysis of whole system is carried out and discussed along with various operational stages of HIES.

Composition design and preparation of lithium lead titanate (Li2PbxTi1-xO3, 0.1 < x < 0.9): A novel tritium breeding ceramic

In deuterium–tritium (D-T) fusion reactors, the design of tritium breeding materials plays very critical role in tritium self-sufficiency. Compared to materials like Li2O, Li2TiO3 has a weaker tritium breeding capability, because of the lower density of Li. For this reason, there is a need to further explore new lithium-containing ceramics. In this paper we have designed a new tritium-breeding Li2PbxTi1-xO3. Tritium production calculations are performed on a 2 m thickness spherical shells of Li2PbxTi1-xO3 with four types of 6Li abundance (7.5 %, 20 %, 50 %, 90 %) and six species of neutron source energy（0.025 eV, 1 MeV, 2.5 MeV, 2.7 MeV, 7.4 MeV and 14.1 MeV）located at the center. It is shown that such a modification may result in the tritium breeding ratio increase to 1.4 in the sphere. Based on simulations, the Li2Pb0.5Ti0.5O3 was synthesized via the sol–gel route, and the microscopic morphology of Li2Pb0.5Ti0.5O3 were briefly investigated. Upon analysis, it's found that lead is uniformly present in the ceramic material in the form of Li2PbO3. In comparison with Li2TiO3 ceramics, Li2PbxTi1-xO3 exhibits stronger tritium breeding capability.

Development of thermo-physical property database and code of fusion materials for CFETR blankets

The conceptual design of China Fusion Engineering Test Reactor (CFETR) has been completed and is currently in the engineering design phase. In this paper, focusing on three categories of candidate blanket designs and divertor designs, the selection schemes for materials of plasma-facing components, solid structure, tritium breeding materials, neutron multiplier, coolant, and divertor were extracted. Thermo-physical property data was compiled to form a material thermo-physical property database, named Thermo-physical Property Database (ThePD), which is the first advanced thermo-physical database for materials of CFETR blankets. Besides, the callable code was generated, which can be directly applied in the system code for the blanket. Based on internationally available database from National Institute of Standards and Technology (NIST) and similar software AP1700, the correctness of the data and the transitional analysis of regions designed by International Association for the Properties of Water and Steam (IAPWS) were verified. Using the thermal-hydraulic system analysis code, integrated validation was performed for the validated thermo-physical property data and code, demonstrating the applicability and accuracy of the data. ThePD is conducive to addressing the issue of incomplete thermo-physical property data faced by blanket design and safety analysis software for CFETR, and it can fill this gap.

The Influence of High-Temperature Tests on the Resistance to Degradation and Reduction in Strength Properties of Lithium-Containing Ceramics Used as Blanket Materials for Tritium Breeding

Conducting high-temperature tests on ceramics-containing lithium, which are employed as tritium breeding materials, plays a crucial role in comprehending their ability to withstand degradation and maintain their strength properties throughout operation. From the standpoint of fusion research, it is imperative to grasp these phenomena in order to guarantee the safety and effectiveness of reactors. Additionally, these factors could impact the choice of particular materials and designs for blanket materials. The primary objective of this research is to evaluate alterations in the strength characteristics of ceramics-containing lithium when subjected to high-temperature thermal stability tests, while also preserving the hardness stability and resistance to cracking in ceramics subjected to cyclic tests. Lithium-containing ceramics based on lithium titanate (Li2TiO3), lithium orthosilicate (Li4SiO4), and lithium methacyrconate (Li2ZrO3), having a high structural ordering degree and good strength properties, were chosen as objects for assessing resistance to high-temperature degradation. During the studies, it was discovered that the presence of interphase boundaries in the composition of ceramics linked to the development of impurity phases results in crack resistance growth during long-term high-temperature tests simulating the stress effect on the material. At the same time, an assessment of high-temperature aging as a result of modeling destruction processes showed that ceramics based on lithium metazirconate are the most resistant to degradation of strength properties. By simulating high-temperature aging processes, it became feasible to establish connections between structural alterations resulting from the thermal expansion of the crystal lattice and oxygen migration phenomena occurring at elevated temperatures. These factors collectively contribute to a detrimental reduction in the strength properties of ceramics-containing lithium.

High-Temperature Intrinsic Defect Chemistry of Li8PbO6 Ceramic Breeding Material.

Understanding the intrinsic defect chemistry of tritium breeder materials proposed for use in future fusion reactors is imperative, as certain defects may act as traps leading to retention of tritium in the ceramic matrix. In this paper, we use combined density functional theory simulations with simple thermodynamics to explore the intrinsic defect chemistry of octalithium plumbate (Li8PbO6) as a function of both temperature and oxygen partial pressure. Importantly, we consider vibrational contributions to the energies of the reference states used in the calculations of the defect formation energies. Our results indicate that including these temperature effects can modify the predicted defect chemistry for materials at a high temperature. For Li8PbO6, the defect chemistry is predicted to be dominated by the VLi-1 defect, which will likely act as a trap for tritium. The charge compensating mechanism is predicted to change as a function of the conditions, with the Lii+1 interstitial defect providing compensation at low temperatures and the VO2+ vacancy defect occurring close to the Li2O saturation limit.

Thermal expansion studies of Li2TiO3 by dilatometry and in-situ high-temperature X-ray diffraction

Li2TiO3 (lithium titanate) is a material of interest in the field of fusion reactors due to its potential application as a tritium breeder material. Because of the extreme working conditions and temperature gradients involved, thermal expansion properties are critical to consider when assessing a material for a fusion reactor. The reactor is subjected to severe neutron flux, high temperature and thermal cycling, which can cause significant stress on the material. A low coefficient of thermal expansion (CTE) of Li2TiO3 is important as it mitigates mechanical stress and minimized the risk of structural damage resulting from thermal cycling. Material degradation, cracking, or delamination might result from a mismatch in CTE between different materials in the reactor. Therefore the estimations of material's thermal expansion properties hold significance for the designing and fabrication of structural material. In this study, Li2TiO3 material was compacted into pellets and high-temperature X-ray diffraction (HT-XRD) was employed to obtain X-ray diffraction patterns from room temperature to 1273 K at 100 K intervals. The lattice parameters (a, b, and c) of Li2TiO3 polycrystals were determined and corresponding unit cell volumes were estimated at each temperature. The thermal expansion percentages for the lattice parameters a, b, and c between room temperature and 1273 K were found to be 1.21%, 1.84%, and 1.83% respectively. The calculated coefficient of thermal expansion (CTE) were determined as 12.5 × 10−6 K−1, 19 × 10−6 K−1, and 18.9 × 10−6 K−1for lattice parameters a, b and c respectively. The crystallographic density at room temperature was estimated to be 3.42 g/cm3. The molar volume was calculated to range between 32.12 and 33.72 cm3/mol from room temperature to 1273 K. Additionally, high-temperature dilatometry was employed to measure the thermal expansion of Li2TiO3, resulting in the estimation of coefficient of thermal expansion. From room temperature to 1290 K, the thermal expansion percentage and mean coefficient of thermal expansion were calculated to be 1.91% and 19.64 × 10−6 1/K, respectively. Furthermore, the temperature dependent mean and instantaneous thermal expansion values were also determined. The differential scanning calorimetry (DSC) method was used to estimate the temperature of the monoclinic to cubic transformation, which was determined to be 1416 K. This paper provides a comprehensive description of the conducted experiments, accompanied by the detailed analysis of the thermal expansion properties derived through the use of High-Temperature X-Ray Diffraction (HT-XRD) and HT-dilatometry (High-Temperature Dilatometry).

Grain growth mechanism and thermal stability of Li2TiO3 pebbles fabricated by wet method

Li2TiO3 ceramic pebbles with uniform structure, high density and good mechanical property were successfully fabricated by wet process. The Li2TiO3 pebbles sintered at 1000 °C for 4 h had satisfactory density (91% T.D.) and crush load (a.v. 27.1 N). The apparent activation energy of Li2TiO3 grain growth was 201 ± 7 kJ/mol. Accelerated densification was observed in a reducing atmosphere, this might be attributed to the oxygen vacancies generated by the reduction of Ti4+ to Ti3+ in 0.1%H2+He atmosphere, which facilitate the diffusion of ions through the lattice. The grain growth was controlled by lattice diffusion under reducing atmosphere, and the apparent activation energy was 168 ± 6 kJ/mol. The mechanical results show that the crush load of Li2TiO3 pebbles decreased with the increase of temperature, and the average crush load at 500 °C was 13.9 N. In addition, the thermal cycling and long-time annealing experiments suggested that Li2TiO3 pebbles have good structural and performance stability. The Li2TiO3 ceramics prepared by wet process were satisfactory tritium breeder materials with ideal application potential.

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.

Effect of the various tritium breeding materials on the tritium breeding ratio in ARC fusion reactor

This study deals with TBR (Tritium Breeding Ratio) performance of the four different types of molten salt (FLiBe, FLiNaK, FLiNaBe, FLiNaRb) which has four different Li-6 enrichment ratios (0%, 30%, 60%, 90%) Moreover, two different vacuum vessel materials (SiC composite and V4Cr4Ti alloy) have been used in ARC reactor geometry. In models with TBR ≥1.05, it has been desired to examine the sustainability of the tritium production of the reactor by removing the neutron multiplication layer from the design without using the toxic Beryllium element. The neutronic performance studies have been examined in OpenMC, which is an open-source Monte Carlo neutron and photon transport code. The TBR values of the different models obtained and the change of these values according to Li-6 enrichment were examined. Obtaining result showed that the FLiBe molten salt and V4Cr4Ti alloy vacuum vessel wall material could be the best choice for the ARC reactor.

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.

A DFT + U study on diffusion and aggregation behavior of He atoms in Li2TiO3

Future fusion power plants are designed based on the deuterium-tritium fusion reactions. Yet, as an element with a half-life of 12.5 years, tritium cannot be obtained in large quantities from nature. To meet future demand, an artificial nuclear reaction is needed to use tritium breeder materials to produce tritium. At present, Li2TiO3 is the most promising tritium breeder blanket material in fusion reactor. The existing experimental results show that helium irradiation will change the crystal structure of Li2TiO3 and cause embrittlement failure. In order to study the microscopic helium embrittlement mechanism of Li2TiO3, the diffusion and aggregation behavior of helium atoms in Li2TiO3 bulk was calculated by first principles. The reasons for the formation of the most stable interstitial configuration of helium atoms and helium clusters in Li2TiO3 bulk are explained by lattice vibrational theory and electronic structure analysis. At the same time, the relatively new research results in this work are mainly as follows: In Li2TiO3 bulk, the three spatial inversion symmetries Li vacancy helium atoms tend to diffuse from all Li vacancies to Li2 vacancies. Meanwhile, the Li2 vacancy is the most likely site for helium clusters to nucleate. And helium atoms and the helium clusters tend to arrange and grow along the plane of the Li layer.

Parametric Study of the Vacuum Permeator for the Tritium Extraction eXperiment

Tritium breeding is a critical component of any self-sustaining future fusion reactor. The liquid-metal eutectic PbLi is of particular interest as a tritium breeder material due to its favorable thermophysical and neutronic properties. One of the several remaining challenges facing PbLi breeder blankets is the need to design and validate a highly efficient tritium extraction system. The vacuum permeator is a promising extraction concept that utilizes tritium permeation through a highly permeable metal membrane. The Tritium Extraction eXperiment (TEX) is a forced-convection PbLi loop constructed to investigate tritium extraction from PbLi with vacuum permeators. Accurate thermal-hydraulic and tritium transport models are required to establish appropriate test matrices, predict experiment outcomes, and analyze data. However, the hydrogen transport properties of PbLi and permeator materials have large uncertainties. A database is collected and a parametric analysis is conducted on the effect of hydrogen transport material properties, including diffusivity of H in PbLi and the permeator, solubility of H in PbLi and the permeator, and the permeator surface recombination constant, on the expected tritium extraction efficiency for a vacuum permeator installed in TEX. Herein, we observe that the solubility of H in PbLi and the permeator and the recombination constant of the permeator have the largest effect on extraction efficiency.

Ab Initio Simulations of Tritium Diffusion in Al-Rich Iron Aluminide Coating Phases

Surface coatings of steels used in extreme conditions and corrosive environments generally aim to provide protection and increased durability. In the case of tritium-producing burnable absorber rods (TPBARs) used in nuclear reactors, a 316 stainless steel has been coated with Al. Scanning transmission electron microscopy (STEM) characterization of the coating found three Al-rich (>60 atom % Al) iron aluminide alloys identified as hexagonal FeNiAl5, monoclinic Fe4Al13, and orthorhombic Fe2Al5. Density functional theory simulations using nudged elastic band have been performed to investigate the diffusion of interstitial tritium in each Al-rich iron aluminide phase. While FeNiAl5 and Fe4Al13 can be viewed as the stacking of two layers, the structural peculiarity of Fe2Al5 is that channels of variable Al vacancy content are present along the c-axis. Therefore, three stoichiometries for Fe2Alx phase, namely, Fe2Al4, Fe2Al5, and Fe2Al6, have been considered to evaluate the impact of Al vacancy concentration on tritium diffusion behavior. Altogether, we found that at 600 K, tritium diffusion decreases from a faster rate in the channels of Fe2Alx phases (DT ≤ 10–11 m2·s–1) to Fe4Al13 (DT ≈ 10–12 m2·s–1), and finally in FeNiAl5 (DT ≈ 10–13 m2·s–1). We also find that interstitial tritium generally diffuses faster in Fe–Al coating phases than in the tritium breeding material γ-LiAlO2 (DT ≈ 10–14 m2·s–1) but slightly slower than in 316 stainless steel (DT ≈ 10–10 m2·s–1).

Li vaporization behavior and mechanical properties of Li8ZrO6 under high temperature reducing atmosphere

Li8ZrO6 sintered compacts are one of the promising candidates for innovative tritium breeding materials from the viewpoint of lithium density, but there are concerns such as a decrease in lithium atomic density and mechanical strength due to Li2O vaporization. In this study, we fabricated a two-phase mixed sintered body of Li8ZrO6 with a high lithium atomic density, as the main component, and relatively a large amount of Li2O, and thermally annealed it in an environment simulating actual use conditions to vaporize lithium. It is shown that this material is stable at 320°C, which is the main temperature in the blanket space, Li2O vaporizes and part of Li8ZrO6 changes to Li6Zr2O7 at least above 700°C, and vaporization of Li2O occurs more rapidly to finally become to a single phase of Li2ZrO3 at 800−900°C. In addition, it is shown that this material has a higher lithium density than Li2TiO3, one of the Japanese candidate materials, in a wide temperature range below 770 °C during the expected period of practical use of three years. Finally, it is concluded that loading Li8ZrO6-based material and Li2TiO3 in the high and low temperature regions, respectively, achieves the highest lithium density.