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JAEA Reports

Calculation by PHITS code for recoil tritium release rate from beryllium under neutron irradiation (Joint research)

Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*

JAEA-Technology 2016-022, 35 Pages, 2016/10

JAEA-Technology-2016-022.pdf:3.73MB

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 $^{4}$ (cm $^{-3}$ ) 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.

JAEA Reports

Study of origin on tritium release into primary coolant for research and testing reactors; Tritium release rate evaluated from JMTR, JRR-3M and JRR-4 operation data

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

JAEA-Technology-2014-025.pdf:43.46MB

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.

Journal Articles

Study of tritium and helium release from irradiated lithium ceramics Li $_{2}$ TiO $_{3}$

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

https://doi.org/10.13182/FST11-A12616

Times Cited Count：13 Percentile：69.64(Nuclear Science & Technology)

no abstracts in English

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

https://doi.org/10.1016/j.jnucmat.2010.12.133

Times Cited Count：16 Percentile：75.8(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

https://doi.org/10.1016/j.fusengdes.2007.05.064

Times Cited Count：25 Percentile：83.32(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

https://doi.org/10.1016/j.fusengdes.2005.07.019

Times Cited Count：41 Percentile：92.46(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 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

Evaluation of curve for tritium release rate into primary coolant for research and testing reactors

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 $^{6}$ Li by (n,) reaction from $^{9}$ 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.

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Calculation by PHITS code for recoil tritium release rate from beryllium under neutron irradiation (Joint research)

Study of origin on tritium release into primary coolant for research and testing reactors; Tritium release rate evaluated from JMTR, JRR-3M and JRR-4 operation data

Study of tritium and helium release from irradiated lithium ceramics Li $_{2}$ TiO $_{3}$

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

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

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

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

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

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

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

Evaluation of curve for tritium release rate into primary coolant for research and testing reactors