Author's personal copy Recovery and recycling of lithium value from spent lithium titanate (Li 2 TiO 3 ) pebbles (original) (raw)
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Recovery and recycling of lithium value from spent lithium titanate (Li 2 TiO 3 ) pebbles
This article appeared in a journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and education use, including for instruction at the authors institution and sharing with colleagues. Other uses, including reproduction and distribution, or selling or licensing copies, or posting to personal, institutional or third party websites are prohibited. In most cases authors are permitted to post their version of the article (e.g. in Word or Tex form) to their personal website or institutional repository. Authors requiring further information regarding Elsevier's archiving and manuscript policies are encouraged to visit: http://www.elsevier.com/authorsrights g r a p h i c a l a b s t r a c t Effects of various process parameters on the recovery of Li-from spent Li 2 TiO 3 pebbles were investigated. From the experimental results it was observed that the leaching rate increases with speed of stirring till 450 rpm and then above 450 rpm; the increase in speed of stirring does not have any significant effect on the leaching rate as shown in the following figure. Effects of other parameters on the Li-recovery from spent Li 2 TiO 3 pebbles are discussed in this paper. a b s t r a c t In the first generation fusion reactors the fusion of deuterium (D) and tritium (T) is considered to produce energy to meet the future energy demand. Deuterium is available in nature whereas, tritium is not. Lithium -6 (Li 6) isotope has the ability to produce tritium in the n, a nuclear reaction with neutrons. Thus lithium based ceramics enriched by Li 6 isotope are considered for the tritium generation for its use in future fusion reactors. Lithium titanate is one such Li-based ceramic material being considered for its some attractive properties viz., high thermal and chemical stability, high thermal conductivity, and low tritium solubility. It is reported in the literature, that the burn up of these pebbles in the fusion reactor will be limited to only 15–17 atomic percentage. At the end of life, the pebbles will contain more than 45% unused Li 6 isotope. Due to the high cost of enriched Li 6 and the waste disposal considerations, it is necessary to recover the unused Li from the spent lithium titanate pebbles. Till date, only the feasibilities of different processes are reported, but no process details are available. Experiments were carried out for the recovery of Li from simulated Li 2 TiO 3 pebbles and to reuse of lithium in lithium titanate pebble fabrication. The details of the experiments and results are discussed in this paper.
Fusion Engineering and Design, 2012
For the development of TBM for fusion reactors, lithium containing ceramics as against the metal are preferred as tritium breeding material. Lithium titanate (Li 2 TiO 3 ) is one such chosen ceramic tritium breeder. Li 2 TiO 3 pebbles are conventionally prepared by sol-gel process and wet process. Solid state reaction of lithium carbonate with titanium dioxide is preferred route for the bulk production of Li 2 TiO 3. Thermo-gravimetric and differential thermal analysis (TG-DTA) techniques have been used in the present study to understand the solid state reaction of intimate mixture of lithium carbonate and titanium dioxide. It was found out that single phase lithium titanate (Li 2 TiO 3 ) is produced at 750 • C and the reaction is completed in 6 h. Fine powders of lithium titanate obtained after milling and classification were mixed with aqueous solution of PVA to prepare green pebbles of desired size and shape. The pebbles were subsequently sintered at 900 • C and the effect of sintering time on the properties of sintered pebbles was studied. The reaction mechanisms and the product qualities obtained by the solid state reaction, extrusion and spherodization techniques are discussed in this paper.
Fusion Engineering and Design, 2007
The irradiation programme EXOTIC (extraction of tritium in ceramics) is carried out within the European framework for the development of the helium cooled pebble bed concept. The EXOTIC-9/1 is the latest experiment in the series of EXOTICs that are irradiated in the high flux reactor in Petten. Tritium release and inventory in lithium containing ceramic pebbles are key properties to be tested in a TBM. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The objective of the EXOTIC-9/1 is to study in-pile tritium release behaviour of the latest developed lithiummetatitanate pebbles (Li 2 TiO 3). The pebbles are produced by a extrusion-spheroidisation-sintering process at CEA. The new pebbles differ with respect to porosity from the lithiummetatitanate ceramics tested in the previous EXOTIC 8 programme. The pebbles have diameter in the range from 0.6 to 0.8 mm. Irradiation of EXOTC-9/1 started at 24 March 2005, and will continue until the end of 2006, in total about 400 irradiation days. The temperature is varied between 340 and 580 • C. Begin of Life (BOL) tritium production rate is 0.56 mCi/min. Based upon the in-pile tritium release measurements and the analysis of the tritium residence time it can be concluded that tritium release in the new batch of the high density Li 2 TiO 3 pebbles irradiated in EXOTIC 9/1 is rather slow compared to the ceramics irradiated in the EXOTIC 8 irradiation campaign. In this paper, the in-pile tritium behaviour will be reported during normal operation and during transients in temperature, purge gas chemistry and gasflow. The collected data is compared to tritium release data from ceramics irradiated in previous EXOTIC experiments with respect to tritium inventory, residence time and porosity.
Journal of Radioanalytical and Nuclear Chemistry, 2017
Lithium titanate and lithium aluminate, proposed as tritium breeding blanket materials in D-T fusion reactor, were analyzed by nuclear analytical techniques as a part of chemical quality control exercise. Concentrations of Li, Ti and Al were quantified by particle induced gamma-ray emission using 4 MeV proton beam from folded tandem ion accelerator facility. Trace elemental concentrations belonging to long-lived nuclides were determined by instrumental neutron activation analysis using neutron flux from Dhruva research reactor at BARC. The analysis results are useful in designing a shielded facility for tritium release studies of these irradiated breeder materials.
Ceramics for fusion reactors: The role of the lithium orthosilicate as breeder
Physica B: Condensed Matter, 2012
Lithium-based oxide ceramics are studied as breeder blanket materials for the controlled thermonuclear reactors (CTR). Lithium orthosilicate (Li 4 SiO 4) is one of the most promising candidates because of its lithium concentration (0.54 g/cm 3), its high melting temperature (1523 K) and its excellent tritium release behavior. It is reported that the diffusion of tritium is closely related to that of lithium, so it is possible to find an indirect measure of the trend of tritium studying the diffusivity of Li þ. In the present work, the synthesis of the Li 4 SiO 4 is carried out by Spray drying followed by pyrolysis. The study of the Li þ ion diffusion on the sintered bodies, is investigated by means of electrical conductivity measurements. The effect of the gray irradiation is evaluated by the impedance spectroscopy method (EIS) from room temperature to 1173 K. The results indicate that the síntesis process employed can produce Li 4 SiO 4 in the form of pebbles, finally the best ion species for the electrical conduction is the Li þ and is shown that the g-irradiation to a dose of 5MGy, facilitate its mobility through the creation of defects, without change in its conduction process.
Li ceramic pebbles chemical compatibility with Eurofer samples in fusion relevant conditions
Journal of Nuclear Materials, 2004
Information on the chemical compatibility between Li ceramic breeders and reactor structural materials is an important issue for fusion reactor technology. In this work, Eurofer samples were placed inside a Li ceramic pebble bed and kept at 600°C under a reducing atmosphere obtained by the flow of a purging gas (He + 0.1vol.%H 2). Titanate and orthosilicate Li pebble beds were used in the experiments and exposure time ranged from 50 to 2000 h. Surface chemical reactions were investigated with nuclear microprobe techniques. The orthosilicate pebbles present chemical reactions even with the gas mixture, whereas for the samples in close contact with Eurofer there is evidence of Eurofer elemental diffusion into the pebbles and the formation of different types of compounds. Although the titanate pebbles used in the chemical compatibility experiments present surface alterations with increasing surface irregularities along the annealing time, there is no clear indication of Eurofer constituents diffusion.
Use of yttrium for the recovery of tritium from lithium at low concentrations
1979
A particularly difficult problem for first generation fusion power reactor systems is the recovery of tritium at very low con centration levels from the reactor blanket when lithium metal is the principal tritium breeding medium. On the basis of recent data and reasonable extrapolations, we show that it is likely that yttrium metal can be used to extract tritium from lithium at concentrations as low as 10" atom fraction tritium in lithium under conditions that are practicable for commercial power machines. This report was prepircd ti in iccoum of work (pcniorrd by the United Siatet Government. Neither (he Lfnlitd Stiiei nor the Uniied Sunt Depuutxnt af tnctgy, nor iny of thdi employee!, not tny of their contncfofi, mbcantnclon. or their employee*, mikrt my wuranry, expreu or implied, tt ucumei my kpl ttifailily or rapamlblliiy (a the jeeuncy, cample ui>eu or uwfulneiiofany InronTHiion, ippintui, produci j/1 prtxxs duelaied, or reprwenb thil la u« wuld not infringe primely owned rffhu. I ••• '•ruiou-fiu.ij a^siiaia i.xt.i.wn
6 Li produces tritium by (n,) nuclear reaction, 6 Li + 1 n → 4 He + 3 H. Lithium titanate (Li 2 TiO 3) enriched with 6 Li, is the most promising candidate for solid test blanket module (TBM) material for fusion reactors. Various processes are reported in literature for the fabrication of Li 2 TiO 3 pebbles for its use as TBM material. A process has been developed based on the solid state reaction of lithium-carbonate and titanium-dioxide for the synthesis of lithium titanate and pebble fabrication by extrusion, spherodiza-tion and sintering. This paper discusses the sequence of steps followed in this process and the properties obtained. Analytical grade titanium-dioxide and lithium-carbonate were taken in stoichiometric ratio and were milled to ensure thorough intimate mixing and obtain fine particles less than 45 m before its calcination at 900 • C. Following calcination, the agglomerated product was again milled to fine particles of size less than 45 m. Aqueous solution of ploy-vinyl-alcohol was added as binder prior to its feeding in the extruder. The extruded strips were spherodized and spherical pebbles were dried and sintered at 900 • C for different duration. Pebbles of desired density and porosity were obtained by suitable combination of sintering temperature and duration of sintering. Properties of the prepared pebbles were also characterized for sphericity, pore size distribution, grain size, crushing load strength, etc. The values were found to be conforming to the desired properties for use as solid breeder. The attractive feature of this process is almost no waste generation.
Journal of Nuclear Materials, 1994
Within the frame of the COMPLIMENT experiment, y-LiAlO, specimens with different microstructures (grain size distributions) were tested in the same environmental conditions to compare the effects caused by 6Li(n, cu)T reaction and by fast neutron scattering, the damaging dose being held at about the same level (1.6-1.8 dpa). The tritium retention times were obtained by the tritium removal of isothermal annealing under He+O.l% H, sweeping gas. In spite of the different Li burnups (2.5% and 0.25%) and the residual tritium concentrations which were found in the irradiated specimens (4.3 Ci/g and 0.09 Ci/g, respectively, for specimens held at 450°C during the irradiations), the kinetics of tritium removal was not found to be discriminated by the two different irradiations. Moreover, the results were found to agree with those previously obtained by the "in-situ" TEQUILA experiment, performed on the same type of Li ceramics. Hence, the apparent first order desorption mechanism has been confirmed to control the kinetics of tritium removal from the porous fine grain y-LiAlO, ceramics.