Lithium Ceramics Research Articles

In-situ observation of surface -OH and -OD on various lithium ceramics

Stretching vibration of OD − on the surface of various lithium ceramics, Li 2SiO 3, Li 4SiO 4, LiAlO 2, Li 2ZrO 3, Li 8ZrO 6 and Li 6Zr 2O 7 were measured at 400°C under controlled D 2O vapor pressure by a diffuse-reflectance FT-IR (Fourier transform infrared) spectroscopy. Broad peak attributed to hydrogen-bonded OD − and sharp one attributed to isolated OD − on the surface were observed. Frequency of the stretching vibration of the isolated OD − was found to have a strong correlation with the Li/O ratio of lithium ceramics.

In-Situ Observation of Surface Hydroxyl Group on Lithium Ceramics at High Temperature by Infrared Absorption Spectroscopy

Hydroxyl groups on the surface of Li 2 O were studied by using a diffuse reflectance method with Fourier transform infrared spectroscopy at high temperature up to 833K under controlled D2O or D 2 partial pressure. Multiple peaks due to the stretching vibration of O-D were observed at 2748, 2717, 2660, 2620, 2540, 2520, 2490cm -1 showing a different behavior with temperature or atmosphere. When D 2 was added to the sweep gas, peak intensity of 2660 and 2620cm -1 decreased. This result indicates that the surface nature is influenced by oxygen potential. Assignment of observed peaks to the surface nature was also discussed.

A Review of Tritium Release Modelling from Lithium Ceramics

Since lithium-containing ceramics (e.g. Li 2 O, LiAlO 2 , Li 4 SiO 4 , Li 2 ZrO 3 , Li 2 TiO 3 ) are considered as breeding materials in the blanket of the next generation fusion reactors, several studies are in progress to evaluate their behaviour under irradiation in both operating and accidental conditions. Based on safety and economic considerations tritium inventory and release are the most critical issues for blanket concept. Investigation of tritium transport processes by using comprehensive physical-mathematical models is one of the current activities in this area. Although some analytical models and numerical methods dealing with tritium transport and release in fine-grained ceramic were already developed and applied to interpret results from in-situ and/or post-irradiation annealing experiments, it is necessary that presently available computer codes enlarge their range of applicability to be able to predict, with increased accuracy, the tritium release response for a wider range of experimental conditions and material characteristics. This paper reviews the tritium modelling activity and summarizes the existing transport models and computer codes highlighting models development and focusing on major changes and evolutionary improvements. 1 Validation of models by comparison of calculated results with experimental ones is also reported and discussed. Areas of future applications are identified and emphasis is placed upon the growing need of developing more accurate computer codes with the aim to improve the accuracy of blanket tritium inventory estimations.

Surface Processes in Tritium Release from Lithium Ceramics

Surface Processes in Tritium Release from Lithium Ceramics

Modeling of surface reaction in tritium release from lithium ceramics and its comparison with transient experiments

Modeling of surface reaction in tritium release from lithium ceramics and its comparison with transient experiments

Modeling of surface reaction in tritium release from lithium ceramics and its comparison with transient experiments

Abstract Among the various tritium transport processes in lithium ceramics, the importance and the detailed mechanism of surface reactions still remains to be elucidated. A dynamic adsorption and desorption model for tritium desorption from lithium ceramics, especially Li2O was constructed. This model was improved in order to simulate tritium release behavior during temperature and sweep-gas composition changes. The surface heterogeneity of HTO release was also included by introducing multiple activation energies for desorption, which are assumed to be related to the number of oxygen vacancies surrounding noted OH− and OT− on the surface. The kinetic coefficients required for calculation were also improved by using recent experimental and/or reported data. Calculation results for sweep-gas composition and temperature changes were compared with BEATRIX-II experiments. The tritium release transient behavior and the chemical form of the released tritium were discussed.

Ceramic pebble bed development for fusion blankets

Research on lithium ceramic breeders has been intensive since the late 1970s. The bulk material properties of several candidate lithium ceramics are essentially available, although there is still much work to be done on properties under irradiation and on the overall behavior in blanket modules. Based on these results, lithium ceramic breeder blankets have been selected in many fusion reactor design studies. These lithium ceramics are incorporated into blankets typically as monolithic pellets or packed pebble beds. There is substantial industrial experience with pebble beds made from other ceramics as catalyst supports, and in fabrication and testing of pebbles for advanced fission reactor fuels. In fusion blankets, the pebble bed form offers several attractive features, including simpler assembly into complex geometries, a uniform pore network and low sensitivity to cracking or irradiation damage. Ceramic breeder pebbles have been a focus for several research groups. In general, the database is similar to that of monolithic pellets for the materials studied; basic production and material property data are available, but the irradiation and engineering database remains sparse. In addition to the basic requirements on any ceramic breeder material (such as low tritium hold-up, compatibility with structure, irradiation stability, etc.), the main pebble bed requirements may be roughly summarized as follows: economic, high yield production rates; high average bed (smear) density; adequate bed thermal conductivity; acceptable purge gas pressure drop; adequate crush strength; tolerance to thermal cycling. In this paper, the international ceramic breeder pebble bed database is reviewed with respect to these pebble bed properties, and the R&D needs for reactor blanket development are assessed.

Characteristics of Li 2O pebbles fabricated by the melting granulation method

Lithium ceramics have been considered as candidates for solid breeder materials for fusion reactors. In lithium ceramics, lithium oxide (Li 2O) is one of the best tritium breeders from the standpoint of high lithium density and high thermal conductivity. Recently, several studies have been carried out on the fabrication of small, spherical Li 2O forms to reduce the induced thermal stress in the breeder. A mass-production process for making small pebbles was developed using the melting granulation method. In the present work, the characteristics of Li 2O pebbles fabricated by this method, and mass transfer properties within a Li 2O pebble bed, are discussed.

Lithium reprocessing technology for ceramic breeders

Lithium ceramics have been receiving considerable attention as tritium breeding materials for fusion reactors. Reprocessing technology development for these materials is proposed to recover lithium, as an effective use of resources and to remove radioactive isotopes. Four potential ceramic breeders (Li 2O, LiAlO 2, Li 2ZrO 3 and Li 4SiO 4) were prepared in order to estimate their dissolution properties in water and various acids (HCl, HNO 3, H 2SO 4, HF and aqua regia). The dissolution rates were determined by comparing the weight of the residue with that of the starting powder (the weight method). Recovery properties of lithium were examined by the precipitation method.

In situ tritium recovery from Li 2O irradiated in a fast neutron flux: BEATRIX-II, Phase II temperature-change canister

BEATRIX-II, Phase II was an in situ tritium recovery experiment designed to characterize the behavior of lithium ceramics irradiated to high burnup in a fast neutron flux. A thin-walled Li 2O ring specimen was irradiated at temperatures in the range from 510 to 640°C in helium sweep gases containing 0, 0.01 and 0.1% H 2 to a 5% lithium burnup. The tritium recovery behavior of Li 2O was characterized by using temperature transients and sweep gas composition changes to effect changes in the tritium inventory.

The VOM/JRR-2 experiments; performance of in-situ tritium release from the lithium ceramics

In-situ tritium release experiments on lithium ceramics used as tritium breeding materials have been carried out in Japan Research Reactor 2 (JRR-2) to support fusion reactor design activity. The in-situ tritium measurement system was specifically designed for the VOM experiment and several techniques in ceramic electrolysis cell, ionization chamber, capsule and associated components were utilized. The knowledge and experience gained from these experiments have been very useful for the design and fabrication of the IEA collaborative irradiation experiment, BEATRIX-II. This report compares the tritium release behavior between single crystal, ring monolithic and sintered pebble of Li 2O in VOM-34 and 44 experiments. The tritium release behavior of Li 2ZrO 3, Li 4SiO 4 and Li 2Be 2O 3 have been investigated in VOM-32 and 48 experiments.

Modeling of the surface reaction of tritium release from lithium ceramics

From the results of in-situ tritium release experiments (TTTEx), it was clarified that tritium release from the Li 2O surface is affected by both the swamping effects with H 2 and H 2O, and the oxygen potential near the surface. A model named DAD was constructed which includes these two effects by considering the dependencies of surface coverages of OH −, O 2− and empty sites (oxygen vacancies on O 2− lattice sites and the interstitial sites not occupied by OH − or O 2−) on the partial pressures of H 2 and H 2O in the sweep gas. In this model, desorption of surface tritium as HT is considered by the recombination with surface OH − when they are surrounded with active oxygen vacancies, and by the excharge reaction with H 2 in the gas phase. The result of this model calculation agreed fairly well with the experimental results.

Tritium release from beryllium discs and lithium ceramics irradiated in the SIBELIUS experiment

The SIBELIUS experiment was designed to obtain information on the compatibility between beryllium and ceramics, as well as beryllium and steel, in a neutron environment. This experiment comprised irradiation of eight capsules, seven of which were independently purged with a He/0.1% H 2 gas mixture. Four capsules were used to examine beryllium/ceramic (Li 2O, LiAlO 2, Li 4SiO 4, and Li 2ZrO 3) and beryllium/steel (Types 316L and 1.4914) compacts. Isothermal anneal experiments have been run on representative beryllium and ceramic disks from each of the four capsules at 550 to 850°C in steps of 100°C. The results indicate that tritium release from the beryllium did not exhibit burst release behavior, as previously reported, but rather a progressive release with increasing temperature. Generally, ∼99% of the tritium was released by 850°C. Tritium release from the ceramic discs was quite similar to the behavior shown in other dynamic tritium release experiments on lithium ceramics. The tritium content in beryllium discs adjacent to a steel sample was found to be significantly lower than that found in a beryllium disc adjacent to a ceramic sample. Recoil of tritium from the ceramic into the beryllium appears to be the source of tritium entering the beryllium, probably residing in the beryllium oxide layer.

Postirradiation examination of BEATRIX-II, phase I

BEATRIX-II is an in-situ tritium recovery experiment that was designed to characterize the behavior of lithium ceramics irradiated to high burnup in a fast neutron flux. Postirradiation examination was carried out on the Phase I vented canisters: one containing a Li 2O ring capable of temperature changes and the other a Li 2O solid specimen with a center temperature of 1000°C. The tritium inventory of the ring specimen at 630°C was determined to be in the range from 0.2–0.6 wppm, while for the solid specimen the inventory varied from 1.4 wppm at the surface to 0.06 wppm at the inner surface. Downstream transport of the Li 2O by the sweep gas was determined to be insignificant from analyses of acid rinses of selected canister surfaces. Densification and restructuring of the solid specimen during irradiation resulted in the development of a central annulus. Ceramography was used to characterize the columnar grain structure and the mechanisms involved in its evolution.

Irradiation of lithium zirconate pebble-bed in BEATRIX-II Phase II

BEATRIX-II was an in-situ tritium recovery experiment that was designed to characterize the behavior of lithium ceramics irradiated to a high burnup, and to assess their suitability for use in a fusion reactor blanket. This paper describes the results from the vented canister containing 29.47 g of lithium zirconate spheres packed in a bed 13.2 mm OD, 2.3 mm ID and 103 mm long. The enriched lithium spheres (85% 6Li) were irradiated to a burnup of 5.2% (total lithium) in a steep temperature profile −400°C edge, 1100°C center. The sweep gas was He-O.1% H 2, with systematic tests using alternate compositions: He-0.01% H 2 and pure He (maximum duration 8 days). Tritium recovery decreased slightly at lower H 2 concentrations; for example, the buildup of inventory during a 4-day test in pure He was 0.8 Ci, approximately 6.5% of the tritium generated in the lithium zirconate during that period. The steadiness of the bed central temperature and the tritium release rate, together with low moisture release indicate good performance of the zirconate bed.

Tritium release kinetics of solid lithium ceramics with irradiation defects

For the assessment of tritium release behaviors of lithium oxide under irradiation, a model reaction scheme was developed, which was based on the interactions of irradiation-induced defects with tritium. F 0 centers (oxygen vacancy trapping two electrons) and their partners, O 2 molecules, were considered to be produced under the blanket conditions. The model predictions were compared with the results of an in situ tritium recovery experiments with Li 2O, called BEATRIX-II, and the observed tritium release behaviors were successfully interpreted by the present model. The reactions were discussed, of which the equilibrium constant and rate constant were to be determined for the further development of the model.

Diffusion of lithium-6 isotopes in lithium aluminate ceramics using neutron depth profiling

Abstract Lithium Ceramics offer tremendous potential as a source for the production of tritium ( 3 H) for fusion power reactors. Their successful application will depend to a great extent upon the diffusion properties of the 6 Li within the matrix. Consequently knowledge od 6 Li concentration gradients in the ceramic matrices is an important requirement in the continued development of the technology. In this investigation, the neutron depth profile (NDP) technique has been applied to the study of concentration profiles of 6 Li in lithium aluminate ceramics, doped with 1.8%, 50% and 95% 6 Li isotopic concentrations. Specimen for analysis were prepared at Battelle (PNL) as pellet discs. Samples for diffusion studies were arranged as diffusion couples in the following manner: 1.8% 6 Li discs/85% 6 Li powder. Experiments were performed at the Texas A&M Nuclear Science Center Reactor Building, utilizing 1 MW equivalent thermal neutron fluxes 3 × 10 11 n / m 2 s . The depth probed by the technique is approximately 15 μ.m. Diffusion coefficients are in the range of 2.1 × 10 −12 to 7.0 × 10 −11 m 2 s −1 for 1.8% 6 Li-doped ceramics annealed at 1200 and 1400° C, for 4 to 48-h anneal times.

Tritium Release Reaction from Surface of Neutron Irradiated Lithium Ceramics

The tritium surface reaction in lithium ceramics is investigated by combining in-situ tritium release experiments under neutron irradiation, out-of-pile experiments such as desorption studies, and modeling efforts. It was found that tritium is released from lithium ceramics in the chemical form of HT or HTO depending on the chemical composition of the sweep gas. The surface nature or existing state of surface hydroxyl group is important and was found to be affected by the sweep gas and sample pretreatment. The nature of the surface hydroxyl group also plays an important role in modeling the tritium release. In-situ observation of the surface hydroxyl group by infrared absorption spectroscopy revealed that the surface is heterogeneous for D 2 O adsorption

Tritium transport in lithium ceramics porous media

A random network model has been utilized to analyze the problem of tritium percolation through porous Li ceramic breeders. Local transport in each pore channel is described by a set of convection-diffusion reaction equations. Long range transport is described by a matrix technique. The heterogeneous structure of the porous medium is accounted for via Monte Carlo methods. The model was then applied to an analysis of the approach to steady state tritium release from solid breeders. Diffusion was found to be the dominant transport process. The time constant for the approach to steady state was found to obey a scaling law with respect to the size L of the network. This scaling law was found to closely approximate the scaling law obeyed by conventional diffusion process in a completely homogeneous medium.

Analysis of Tritium Transport Mechanisms at the Surface of Lithium Ceramics

AbstractAn analysis was made of the surface fluxes of hydrogen species in order to determine the activation energies which affect the bulk and surface inventories in a lithium ceramic. It was found that in the absence of protium, Sievert Law is obeyed and the concentration depends on However, in the presence of protium, the concentration depends on PHT. The bulk and surface concentrations were found to depend on a combination of all four surface energies in the activation energy of solution and the heat of adsorption.