Lithium Ceramic Materials Research Articles

Dissociation mechanism of H2 molecule on the Li2O/hydrogenated-Li2O (111) surface from first principles calculations

Hydrogen molecules in a purge gas are known to enhance the release of tritium from lithium ceramic materials, which has been demonstrated in numerous in-pile experiments.

Surface inventory of tritium on Li 2TiO 3

Though lithium ceramic materials such as Li 2O, LiAlO 2 , Li 2 ZrO 3 , Li 4 SiO 4 and Li 2TiO 3 are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials has not been fully understood, yet. It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of breeder material, absorption of water vapor into bulk of the grain, and adsorption of water on surface of the grain, together with two types of isotope exchange reactions for estimation of the tritium inventory in a uniform solid breeder blanket under the steady-state condition. The adsorption capacity and the isotope exchange capacity on the Li 2TiO 3 surface are experimentally obtained in this study. Comparison of the tritium inventory estimated for the Li 2TiO 3 blanket with the inventory for LiAlO 2, Li 2ZrO 3 or Li 4SiO 4 blanket using the model presented by the present authors shows that the Li 2TiO 3 blanket gives the smallest tritium inventory among the candidate ceramic breeding materials.

Isotope exchange reaction on solid breeder materials

Lithium ceramic materials such as Li 2O, LiAlO 2, Li 2ZrO 3, Li 2TiO 3 and Li 4SiO 4 are considered to be as candidate for the tritium breeding material in a deuterium–tritium (D–T) fusion reactor. In the recent blanket designs, helium gas with hydrogen or deuterium is planned to be used as the blanket purge gas to reduce tritium inventory and promote tritium release from the breeding material. In addition, the rate of isotope exchange reaction between hydrogen isotopes in the purge gas and tritium on the surface of the breeding material is necessary to analyze the tritium release behavior from the breeding materials. However, the rate of isotope exchange reactions between hydrogen isotopes in the purge gas and tritium on the surface of those materials has not been quantified until recently. Recently, the present authors quantified the rate of isotope exchange reaction on Li 2O and Li 2ZrO 3. The overall mass transfer coefficients representing the isotope exchange reaction between H 2 and D 2O on breeding materials or the same between D 2 and H 2O are experimentally obtained in this study. Comparison to isotope exchange reaction rates on various breeding materials is also performed in this study. Discussions about the effects of temperature, concentration of hydrogen in the purge gas or flow rate of the purge gas on the conversion of tritiated water to tritium gas are also performed.

Tritium inventory estimation in solid blanket system

Although lithium ceramic materials such as Li 2O, LiAlO 2, Li 2ZrO 3 and Li 4SiO 4 are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials is not yet fully understood. Most results of in situ tritium release experiments are analyzed assuming that the overall release process of tritium is mainly controlled with tritium diffusion in the crystal grain of a solid breeder material. However, the diffusivities by various authors for solid breeder materials do not agree with each other. An insufficient estimation of the surface reactions, the effect due to irradiation defects or the system effect may be the reason. Effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reaction on the tritium inventory in the steady-state condition, are discussed in this study. Comparisons of tritium inventory estimated in this study with data obtained in several in-situ experiments are also performed, and good agreement is obtained when the existence of some water vapor is assumed in the purge gas.

Isotope exchange capacity of solid breeder materials

Though lithium ceramic materials such as Li 2O, LiAlO 2, Li 2ZrO 3 and Li 4SiO 4 are considered as the candidates of tritium breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials is not yet fully understood. It is observed in this study that LiAlO 2 or Li 4SiO 4 has an additional capacity for tritium uptake on the surface through isotope exchange reactions, the isotope exchange capacity, other than the adsorption capacity. It is considered that some -OH bases attributed to the strongly bound chemically adsorbed water or the crystal water formed on the grain surface act as tritium trap, when the isotope exchange reaction are active. It is also observed in this study that Li 2ZrO 3 has no isotope exchange capacity. Effects of the isotope exchange capacity on the tritium inventory in the breeding blanket is discussed based on the data observed in this study where effects of diffusion of tritium in the grain, absorption of water in the bulk of grain and adsorption of water on the surface of grain together with isotope exchange reactions are considered.

Tritium inventory in a LiAlO 2 blanket

Although lithium ceramic materials such as Li 2O, LiA1O 2, Li 2ZrO 3 and Li 4SiO 4 are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials has not been fully understood. Most of the results of in situ tritium release experiments are analyzed assuming that the overall release process of tritium is mainly controlled with the process of tritium diffusion in the crystal grain of a solid breeder material. However, it has been pointed out by many authors that contribution of the surface reaction cannot be ignored. Effects of diffusion of tritium in the grain, absorption of water in the bulk of grain and adsorption of water on the surface of grain together with isotope exchange reactions on the tritium inventory under the steady state condition are discussed in this study. Comparison of the estimated tritium inventory using the way of this study with the data obtained in several in situ experiments shows that good agreements are obtained when the effective diffusivity obtained from the data by Kwast et al. is used. The better agreements are obtained when existence of some water vapor is assumed in the purge gas.

Mistral: A comprehensive model for tritium transport in lithium-base ceramics: Part II: Comparison of model predictions with experimental results

A new tritium transport model called MISTRAL ( M odel for I nvestigative S tudies of T ritium R elease in L ithium Ceramics) has been developed to describe and predict the kinetics of tritium release in lithium ceramic materials for tritium breeding applications in fusion blankets. The model has transient capabilities and has been developed to analyze the full range of transient conditions produced in in-pile tritium recovery experiments and expected in fusion blankets. Calibration of the model against experiments has been done in parallel with its development in order to assess its predictive capabilities and to identify the ranges of potential applicability. The comparisons of the results available for lithium metasilicate and aluminate samples irradiated respectively in the two in-pile tritium recovery experiments LISA1 and MOZART are presented and discussed in this paper. They have been selected for the calibration of the codes as being good examples of various features relevant for tritium release analysis in ceramic breeders under different transient conditions such as change in temperature, purge gas composition and reactor power.

Chemisorption and implantation of hydrogen isotopes in lithium-based fusion blanket ceramics and their thermal release: A study by nuclear reaction analysis

A brief summary is presented on tritium implantation techniques and depth profiling analysis by elastic recoil detection (ERD) and T(d, α)n nuclear reaction analysis (NRA), as applied to the studies of lithium ceramic materials to be used as tritium-breeding blankets in fusion reactors. New data about thermal behaviour of hydrogen and tritium near the surface of both poly- and monocrystalline Li 2O, and ceramic γ-LiAlO 2, are discussed. Ion beam techniques are especially suitable for studying near-surface diffusion and chemisorption of tritium, as well as for modelling the influence of irradiation damage on tritium transport phenomena.

Measurements of tritium and helium in fast neutron irradiated lithium ceramics using high temperature vacuum extraction

Tritium and helium retention measurements have been performed at the Hanford Engineering Development Laboratory on fast-neutron-irradiated lithium ceramic materials from solid-breeder closed-capsule test FUBR-1A. The lithium ceramic materials discussed at this time include Li 2 0, LiAlO 2 , and Li 2 ZrO 3 . The experimental conditions forming the test matrix include 6 Li burnup levels of 1, 2 and 3 at% and irradiation temperatures of 773, 973 and 1173 K, for the three materials. Included will be a discussion of mechanisms controlling tritium retention. The retention measurements reported here were obtained using a rapid, high temperature vacuum extraction method, utilizing newly designed apparatus. This new apparatus will be described, along with developmental supporting data.