Refine your search:     
Report No.
 - 
Search Results: Records 1-8 displayed on this page of 8
  • 1

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Tritium accumulation and release from Li$$_{2}$$TiO$$_{3}$$ during long-term irradiation in the WWR-K reactor

Tazhibayeva, I.*; Beckman, I.*; Shestakov, V.*; Kulsartov, T.*; Chikhray, E.*; Kenzhin, E.*; Kuykabaeva, A.*; Kawamura, Hiroshi; Tsuchiya, Kunihiko

Journal of Nuclear Materials, 417(1-3), p.748 - 752, 2011/10

 Times Cited Count:16 Percentile:75.9(Materials Science, Multidisciplinary)

For the first time the data was obtained on tritium release from $$^{6}$$Li-enriched (96%) lithium metatitanate under high lithium burn-up (up to 23%). Proposed mathematics and software of the reactor experiments allowed to interpret the experimental results of tritium release study. Tritium was continuously generated as a result of the nuclear reaction of lithium-6 and thermal neutrons under variable thermal impacts (graduated heating and cooling) on lithium metatitanate Li$$_{2}$$TiO$$_{3}$$. Main gas release parameters were calculated in order to assess acceptability of the use of lithium metatitanate granules in tritium breeders; the parameters are as follows: gas release rate, tritium retention in the materials, retention time, activation energy of thermal desorption HT, activation energy of volume diffusion T$$^{+}$$, as well as corresponding pre-exponential (frequency) indexes. It was discovered that the tritium release process is mainly controlled by tritium volume diffusion, however, capture of tritium by the point defects and tritium molization at the material's surface played the certain role in the process as well. It was discovered that as lithium is burnt-up, the activation energy of tritium release decreases and tends to a constant value under high lithium-6 burn-up.

Journal Articles

Tritium generation in lithium ceramics Li$$_{2}$$TiO$$_{3}$$ for fusion reactor blanket

Tazhibayeva, I. L.*; Kenzhin, E. A.*; Kulsartov, T. V.*; Kuykabayeva, A. A.*; Shestakov, V.*; Chikhray, E.*; Gizatulin, S.*; Maksimkin, O. P.*; Beckman, I. N.*; Kawamura, Hiroshi; et al.

Questions of Atomic Science and Technology, 2, p.3 - 11, 2008/00

Lithium titanate (Li$$_{2}$$TiO$$_{3}$$) was chosen as a tentative reference material from viewpoints of good tritium recovery at low temperatures and of low tritium inventory and chemical stability for the breeding blanket in fusion reactors. The results of the irradiation tests of Li$$_{2}$$TiO$$_{3}$$ in the WWR-K of NNC-RK are described in this paper. 96at% $$^{6}$$Li-enriched Li$$_{2}$$TiO$$_{3}$$ pebbles and disks were prepared as the irradiation specimens and these specimens were irradiated during 220 days (5350 hours) at the reactor power of 6 MWt. Tritium release was measured continuously during irradiation tests and tritium release properties were evaluated. The mechanics describing generation and release of tritium from the irradiated Li$$_{2}$$TiO$$_{3}$$ were analyzed. There was estimated tritium loss due to recoil energy and binding of tritium in HTO, and there was calculated stationary tritium release due to diffusion under constant temperature and under thermal cycling.

Journal Articles

Structure, composition and properties of lithium ceramic Li$$_{2}$$TiO$$_{3}$$+5% mole TiO$$_{2}$$ irradiated in WWR-K reactor for solid ceramic blanket of fusion reactor

Tazhibayeva, I. L.*; Kulsartov, T.*; Kenzhin, E. A.*; Maksimkin, O. P.*; Doronina, T. A.*; Silnyagina, N. S.*; Turubarova, L. G.*; Tsai, K. V.*; Zheltov, D. A.*; Kashirskiy, V. V.*; et al.

Questions of Atomic Science and Technology; Series the Thermonuclear Fusion, 1, p.3 - 11, 2008/00

The paper contains and analyzes the results of integrated material studies of lithium ceramic Li$$_{2}$$TiO$$_{3}$$ + 5% mole TiO$$_{2}$$ irradiated in reactor WWR-K during 5,350 hours under controlled conditions taking into account effects of tritium generated in the course of irradiation. The changes in density, microstructure, phase and chemical composition, strength and microhardness were studies; lithium burn-up level and tritium residual content were defined. The significant influence of radiation-thermal impacts on structure and properties of ceramic samples were observed. It was shown that irradiation resulted in lithium ceramics softening, at that this effect depended on irradiation temperature. It was discovered the radiation change of phase composition of lithium ceramic.

Journal Articles

In-pile tritium permeation through F82H steel with and without a ceramic coating of Cr$$_{2}$$O$$_{3}$$-SiO$$_{2}$$ Including CrPO$$_{4}$$

Nakamichi, Masaru; Kulsartov, T. V.*; Hayashi, Kimio; Afanasyev, S. E.*; Shestakov, V. P.*; Chikhray, Y. V.*; Kenzhin, E. A.*; Kolbaenkov, A. N.*

Fusion Engineering and Design, 82(15-24), p.2246 - 2251, 2007/10

 Times Cited Count:25 Percentile:83.37(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Use of WWR-K reactor for long-term trials of lithium ceramic Li$$_{2}$$TiO$$_{3}$$ for fusion reactor blanket

Tazhibayeva, I. L.*; Kenzhin, E. A.*; Chachrov, P. V.*; Arinkin, F. M.*; Gasatulin, Sh. Kh.*; Bekamukhabetov, E. S.*; Shestakov, V. P.*; Chikhray, E. V.*; Kulsartov, T. V.*; Kuykabaeva, A. A.*; et al.

Questions of Atomic Science and Technology; Series the Thermonuclear Fusion, 2, p.3 - 10, 2007/00

no abstracts in English

Journal Articles

Investigation of hydrogen isotope permeation through F82H steel with and without a ceramic coating of Cr$$_{2}$$O$$_{3}$$-SiO$$_{2}$$ including CrPO$$_{4}$$, Out-of-pile tests

Kulsartov, T. V.*; Hayashi, Kimio; Nakamichi, Masaru*; Afanasyev, S. E.*; Shestakov, V. P.*; Chikhray, Y. V.*; Kenzhin, E. A.*; Kolbaenkov, A. N.*

Fusion Engineering and Design, 81(1-7), p.701 - 705, 2006/02

 Times Cited Count:40 Percentile:92.5(Nuclear Science & Technology)

no abstracts in English

Oral presentation

Main results of long-term high lithium burn-up irradiation test in Li$$_{2}$$TiO$$_{3}$$ and Li$$_{2}$$TiO$$_{3}$$ + 5mol% TiO$$_{2}$$ ceramics for solid breeding blanket

Tazhibayeva, I.*; Kenzhin, E. A.*; Kulsartov, T.*; Beckman, I.*; Chikhray, E.*; Shestakov, V. P.*; Kuykabaeva, A.*; Maksimkin, O.*; Kawamura, Hiroshi; Tsuchiya, Kunihiko

no journal, , 

The paper contains the results of the integrated material study of lithium ceramics Li$$_{2}$$TiO$$_{3}$$ and Li$$_{2}$$TiO$$_{3}$$ + 5mol% TiO$$_{2}$$ enriched by $$^{6}$$Li (up to 96%). The ceramics were irradiated in the WWR-K reactor during 5350 hours under the temperature range of 400-900$$^{circ}$$С with ${it in situ}$ study of tritium generated during irradiation. The post-radiation studies allowed to determine quantity of residual tritium, degree of lithium burn-up, strength characteristics of lithium ceramic with the lithium burn-up up to 20-23%, ceramic density, changes in the sample microstructure, heat characteristic of the ceramics and their changes due to neutron irradiation, changes of element and phase composition of the samples, and the parameters of tritium release from lithium ceramics. It was showed that the ceramic samples irradiated under lower temperature are characterized by sufficiently small degree of $$^{6}$$Li burn-up. It was established that irradiation resulted in softening of lithium ceramic; at that the effect is more prominent for lower irradiation temperatures. The quantity of tritium released during a reactor's campaign is somewhat increasing with increase of a campaign's number, but quantity of tritium released from lithium titanate per hour doesn't depend on duration of irradiation. Thus, despite of lithium burn-up, tritium flow from lithium titanate isn't changed during long-term irradiation since reduction of the strength of the tritium source (due to lithium burn-up) is compensated by increase in mobility of tritium in defect lattice. The obtained results showed that a breeder on the basis of $$^{6}$$Li-enriched lithium titanate can be a permanent source of tritium during one year of reactor operation at least.

Oral presentation

Collaboration of JAEA and NNC for Kazakhstan project on high-temperature gas-cooled reactor

Nakatsuka, Toru; Levin, A. G.*; Ueta, Shohei; Gizatulin, S.*; Tachibana, Yukio; Kolodeshnikov, A.*; Sakaba, Nariaki; Chakrov, P.*; Kunitomi, Kazuhiko; Vassiliev, Y. S.*; et al.

no journal, , 

The small-sized high-temperature gas-cooled reactors (HTGRs) with an electric power rating of less than 300 MWe can greatly facilitate decentralized energy supply, and create new industries and stimulate economical development in cities and localities as well as in those remote regions to which power transmission grids are undeveloped in developing countries such as Kazakhstan. In 2007, Japan Atomic Energy Agency (JAEA) and National Nuclear Center of Kazakhstan (NNC) have started to collaborate in nuclear energy research and development for early realization of deployment of the HTGR in Kazakhstan, and to support for the Kazakhstan HTGR (KHTR) Project by utilizing the technologies developed under the High Temperature Engineering Test Reactor (HTTR) Project. In 2010, JAEA started a conceptual design of KHTR steam turbine system with thermal power of 50 MW and the maximum coolant temperature at reactor outlet of 750 $$^{circ}$$C for earlier development of HTGRs with support of Japan parties, which consists of Japanese industrial companies, etc. in order to support NNC for preparation of the feasibility study of KHTR.

8 (Records 1-8 displayed on this page)
  • 1