Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Nagata, Hiroshi; Kochiyama, Mami; Chinone, Marina; Sugaya, Naoto; Nishimura, Arashi; Ishikawa, Joji; Sakai, Akihiro; Ide, Hiroshi
JAEA-Data/Code 2024-016, 44 Pages, 2025/03
JAEA-Data-Code-2024-016.pdf:3.54MB
The elemental composition of the structural materials of nuclear reactor facilities is used as one of the important parameters in activation calculations that are evaluated when formulating decommissioning plans. Regarding the elemental composition of aluminum alloys and other materials used as structural materials for test and research reactors, sufficient data is not available regarding elements other than the major elements. For this reason, samples were collected from aluminum alloy, beryllium, hafnium, and other materials that have been used as the main structural materials of JMTR (Japan Materials Testing Reactor), and their elemental compositions were analyzed. This report summarizes the elemental composition data of 78 elements obtained in FY2023.
Nakano, Keita; Matsuda, Hiroki*; Meigo, Shinichiro; Iwamoto, Hiroki; Takeshita, Hayato*; Maekawa, Fujio
JAEA-Research 2021-014, 25 Pages, 2022/03
JAEA-Research-2021-014.pdf:2.1MB
For the development of accelerator-driven transmutation system (ADS), measurement of nuclide production cross-sections in proton-induced reactions on 
Be, C, 
Al, 
Sc, and V have been performed. The measured data are compared with the calculations by the latest nuclear reaction models and with the nuclear data library to investigate the reproducibilities.
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：1 Percentile：9.12(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; Sakamoto, Naoki*
Physical Sciences and Technology, 6(2), p.60 - 63, 2019/12
Tritium release into the primary coolant of the research and test reactors during operation had been studied, and it is found that the recoil release from chain reaction of Be is dominant. To reduce tritium concentration of the primary coolant, feasibility study of the tritium recoil barrier for the beryllium neutron reflectors was carried out, and the tritium recoils of various materials were calculated by PHITS. From these calculation results, it is clear that the thickness of tritium recoil barrier depends on the material and 20
40 
m is required for three orders reduction.
Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
JAEA-Technology 2018-010, 33 Pages, 2018/11
JAEA-Technology-2018-010.pdf:2.58MB
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.
Ishitsuka, Etsuo; Kenzhina, I. E.*
Physical Sciences and Technology, 4(1), p.27 - 33, 2018/06
Increase of tritium concentration in the primary coolant for the research and testing reactors during reactor operation had been reported. To clarify the tritium sources, a curve of the tritium release rate into the primary coolant for the JMTR and the JRR-3M are evaluated. 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. These results show the beryllium components in core strongly affect to the tritium release into the primary coolant. 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 the 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.
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 2
10
(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.
Verzilov, Y. M.; Sato, Satoshi; Ochiai, Kentaro; Wada, Masayuki*; Klix, A.*; Nishitani, Takeo
Fusion Engineering and Design, 82(1), p.1 - 9, 2007/01
Times Cited Count：10 Percentile：56.92(Nuclear Science & Technology)no abstracts in English
Kondo, Keitaro; Takagi, Satoshi*; Murata, Isao*; Miyamaru, Hiroyuki*; Takahashi, Akito*; Kubota, Naoyoshi; Ochiai, Kentaro; Nishitani, Takeo
Fusion Engineering and Design, 81(8-14), p.1527 - 1533, 2006/02
Times Cited Count：14 Percentile：67.04(Nuclear Science & Technology)In a fusion reactor development, double-differential charged-particle emission cross sections(DDXc) are necessary to estimate nuclear heating and material damages of candidate materials irradiated with neutrons. Especially in light nuclei such as Be, Li, and C, nuclear reactions are complicated and difficult to estimate energy spectra of emitted particles based only on theoretical calculations Therefore development of a new technique for precise measurements of DDXc is quite important. Recently we successfully developed a new spectrometer for measurement of DDXc using a pencil-beam DT neutron source of FNS in JAERI. In the present study we carried out measurements of DDXc of Be, 
C, 
F, and Al using the spectrometer. The present technique was valid from the result of measurement for the Al(n,x) reactions. Slight differences appeared between measured data and evaluation or previous experimental values for Be(n,x) reactions at backward scattering angle and in lower energy region.
Imai, Makoto*; Shirai, Toshizo*; Saito, Manabu*; Haruyama, Yoichi*; Ito, Akio*; Imanishi, Nobutsugu*; Fukuzawa, Fumio*; Kubo, Hirotaka
Journal of Plasma and Fusion Research SERIES, Vol.7, p.323 - 326, 2006/00
no abstracts in English
Hegeman, J. B. J.*; Van der Laan, J. G.*; Kawamura, Hiroshi; M
slang, A.*; Kupriyanov, I.*; Uchida, Munenori*; Hayashi, Kimio
Fusion Engineering and Design, 75-79, p.769 - 773, 2005/11
Times Cited Count：27 Percentile：84.19(Nuclear Science & Technology)no abstracts in English
Verzilov, Y. M.; Sato, Satoshi; Nakao, Makoto*; Ochiai, Kentaro; Wada, Masayuki*; Nishitani, Takeo
JAERI-Research 2004-015, 55 Pages, 2004/10
JAERI-Research-2004-015.pdf:3.29MB
no abstracts in English
Suzuki, Satoshi
Koon Gakkai-Shi, 30(5), p.243 - 247, 2004/09
Selection and the development of plasma facing materials for fusion devices, mainly ITER, are presented. For the divertor, CFC (Carbon fiber reinforced carbon composite) materials are utilized as plasma facing materials in the lower part of vertical targets in ITER. Since the design maximum heat flux to the vertical targets is 20 MW/m
, CFC materials, which have higher thermal conductivity than pure copper, are preferable from a heat removal point of view. On the other hand, a plasma facing material of a dome and a liner is tungsten because tungsten has low sputtering yield and has relatively high thermal conductivity among metals. First wall covers 80% of the plasma facing area of ITER. The plasma facing material of the first wall should have good compatibility with plasma. Therefore, beryllium is utilized as a plasma facing material from the low contamination and the minimization of the oxygen impurity to the plasma points of view.
Verzilov, Y. M.; Ochiai, Kentaro; Klix, A.; Sato, Satoshi; Wada, Masayuki*; Yamauchi, Michinori*; Nishitani, Takeo
Journal of Nuclear Materials, 329-333(Part2), p.1337 - 1341, 2004/08
Times Cited Count：4 Percentile：28.74(Materials Science, Multidisciplinary)no abstracts in English
Verzilov, Y. M.; Ochiai, Kentaro; Sato, Satoshi; Wada, Masayuki*; Yamauchi, Michinori*; Nishitani, Takeo
JAERI-Research 2004-005, 30 Pages, 2004/03
JAERI-Research-2004-005.pdf:1.93MB
no abstracts in English
Nishitani, Takeo; Ochiai, Kentaro; Klix, A.; Verzilov, Y. M.; Sato, Satoshi; Yamauchi, Michinori*; Nakao, Makoto*; Hori, Junichi; Enoeda, Mikio
Proceedings of 20th IEEE/NPSS Symposium on Fusion Engineering (SOFE 2003), p.454 - 457, 2003/10
no abstracts in English
Hatano, Toshihisa; Enoeda, Mikio; Kuroda, Toshimasa*; Akiba, Masato
JAERI-Tech 2002-075, 59 Pages, 2002/10
JAERI-Tech-2002-075.pdf:17.38MB
no abstracts in English
Kai, Tetsuya; Teshigawara, Makoto; Watanabe, Noboru; Harada, Masahide; Sakata, Hideaki*; Ikeda, Yujiro
Journal of Nuclear Science and Technology, 39(2), p.120 - 128, 2002/02
Times Cited Count：13 Percentile：62.41(Nuclear Science & Technology)no abstracts in English
Nakamura, Hirofumi; Hayashi, Takumi; Ohira, Shigeru; Nishi, Masataka
JAERI-Research 2001-042, 21 Pages, 2001/09
JAERI-Research-2001-042.pdf:1.82MB
Study on Implantation Driven Permeation (IDP) behavior of deuterium through pure beryllium was investigated as a part of the research to predict the tritium permeation through the first wall components of ITER (International Thermonuclear Experimental Reactor). The permeation experiments were carried out with two beryllium specimens, one was an unannealed specimen and the other was that annealed at 1173 K. The permeation flux was measured as a function of specimen temperature and incident ion flux. Surface analysis of specimen was also carried out after the permeation experiment. Permeation was observed only with the annealed specimen and no significant permeation was observed with unannealed specimen under the present experimental condition (maximum temperature: 685K, detection limit: 1x10
Datoms/ms). It could be attributed that the intrinsic lattice defects, which act as diffusion preventing site, decreased with the specimen annealing. Based on the result of steady and transient permeation behavior and surface analysis, it was estimated that the deuterium permeation implanted into annealed beryllium was controlled by surface recombination due to the oxide layer on the surface of the permeated side.
