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Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.
JAEA-Technology 2021-016, 16 Pages, 2021/09
As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (183 layers) fuel blocks with 20% enrichment of U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.
Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03
Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by Li(n,)H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.
Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a research and testing reactors containing beryllium reflectors.
Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.
JAEA-Technology 2020-008, 16 Pages, 2020/08
As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.
Chikhray, Y.*; Askerbekov, S.*; Kenzhin, Y.*; Gordienko, Y.*; Ishitsuka, Etsuo
Fusion Science and Technology, 76(4), p.494 - 502, 2020/05
Times Cited Count:1 Percentile:12.16(Nuclear Science & Technology)Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.
JAEA-Technology 2019-008, 12 Pages, 2019/07
As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.
Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
JAEA-Technology 2018-010, 33 Pages, 2018/11
As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.
Chikhray, Y.*; Kulsartov, T.*; Shestakov, V.*; Kenzhina, I.*; Askerbekov, S.*; Sumita, Junya; Ueta, Shohei; Shibata, Taiju; Sakaba, Nariaki; Abdullin, Kh.*; et al.
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.572 - 577, 2016/11
Application of SiC as corrosion-resistive coating over graphite remains important task for HTGR. This study presents the results of chemical interaction of the SiC gradient coating over the high-density IG-110 graphite with water vapor in the temperature up to 1673 K. The experiments at 100 Pa of water vapor showed that the passive reaction caused to form SiO film on the surface of SiC coating. Active corrosion of SiC in 1Pa of water vapor leads to deposits of various carbon composites on its surface.
Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
JAEA-Technology 2016-022, 35 Pages, 2016/10
As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, the calculation methods by PHITS code is studied to evaluate the recoil tritium release rate from beryllium core components. Calculations using neutron and triton sources were compared, and it is clear that the tritium release rates in both cases show similar values. However, the calculation speed for the triton source cases is two orders faster than that for the neutron source case. It is also clear that the calculation up to history number per unit volume of 210 (cm) is necessary to determine the recoil tritium release rate of two effective digits precision. Furthermore, the relationship between the beryllium shape and recoil tritium release rate using the triton sources was studied. Recoil tritium release rate showed linear relation to the surface area per volume of beryllium, and the recoil tritium release rate showed about half of the conventional equation value.
Ishitsuka, Etsuo; Motohashi, Jun; Hanawa, Yoshio; Komeda, Masao; Watahiki, Shunsuke; Mukanova, A.*; Kenzhina, I. E.*; Chikhray, Y.*
JAEA-Technology 2014-025, 77 Pages, 2014/08
It has been shown that tritium concentration in the primary coolant of the JMTR and JRR-3M increases during its operation. In this report, to clarify the tritium sources, the tritium release rate into the primary coolant in each operation cycle for the JMTR, JRR-3M and JRR-4 was evaluated. As a result, the tritium release rate is 8 Bq/Wd in the JRR-4, which has not the beryllium core components installed, and no increase in the tritium concentration during reactor operation is observed. In contrast, the tritium release rate is about 1095 and 60140 Bq/Wd in the JRR-3M and JMTR respectively, which cores contain beryllium components, and where the tritium content increases while reactor operates. It is also observed that the amount of released tritium is lower in the case of new beryllium components installation, and increases with the reactor operating cycle.
Kulsartov, T.*; Tazhibayeva, I.*; Gordienko, Y.*; Chikhray, E.*; Tsuchiya, Kunihiko; Kawamura, Hiroshi; Kulsartova, A.*
Fusion Science and Technology, 60(3), p.1139 - 1142, 2011/10
Times Cited Count:13 Percentile:69.64(Nuclear Science & Technology)no abstracts in English
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:75.8(Materials Science, Multidisciplinary)For the first time the data was obtained on tritium release from 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 LiTiO. 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.
Chikhray, Y.*; Shestakov, V.*; Maksimkin, O.*; Turubarova, L.*; Osipov, I.*; Kulsartov, T.*; Kuykabayeba, A.*; Tazhibayeva, I.*; Kawamura, Hiroshi; Tsuchiya, Kunihiko
Journal of Nuclear Materials, 386?388, p.286 - 289, 2009/04
Times Cited Count:19 Percentile:76.21(Materials Science, Multidisciplinary)The PIE (Post Irradiation Examinations) results of LiTiO pebbles added with 5 mol% TiO after the long-term irradiation tests are described in this paper. 96 at% Li-enriched LiTiO pebbles were prepared as the irradiation specimens and these specimens were irradiated during 223 days at the reactor power of 6 MWt in the WWR-K of NNC-RK. After neutron irradiation, light-colored pebbles became gray-colored due to structure changes which generation of gray-colored inclusions with low density and microhardness. Crystal structure of the pebbles after the irradiation test were changed from the results of X-ray diffraction measurement. The value of maximum permissible load (pebble crash limit) at that was also low. The residual tritium in the pebbles was measured after the irradiation test.
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 (LiTiO) 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 LiTiO in the WWR-K of NNC-RK are described in this paper. 96at% Li-enriched LiTiO 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 LiTiO 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.
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 LiTiO + 5% mole TiO 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.
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:83.32(Nuclear Science & Technology)no abstracts in English
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
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:41 Percentile:92.46(Nuclear Science & Technology)no abstracts in English
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 LiTiO and LiTiO + 5mol% TiO enriched by Li (up to 96%). The ceramics were irradiated in the WWR-K reactor during 5350 hours under the temperature range of 400-900С with 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 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 Li-enriched lithium titanate can be a permanent source of tritium during one year of reactor operation at least.
Kenzhina, I. E.*; Ishitsuka, Etsuo; Okumura, Keisuke; Takemoto, Noriyuki; Mukanova, A.*; Chikhray, Y.*
no journal, ,
Increase of tritium concentration in the primary coolant for research and testing reactors during reactor operation had been reported. To clarify the tritium sources, a curve of tritium release rate into the primary coolant for the JMTR and JRR-3M are evaluated. As a result, the tritium release rate is related with produced Li by (n,) reaction from Be, and evaluation results of tritium release curve are shown as the dominant source of tritium release into the primary coolant for the JMTR and JRR-3M are beryllium components. Scattering of the tritium release rate with irradiation time were observed, and this phenomena in the JMTR occurred in earlier time than that of the JRR-3M.