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Yokoyama, Kenji; Ishikawa, Makoto*
Annals of Nuclear Energy, 154, p.108100_1 - 108100_11, 2021/05
Times Cited Count:1 Percentile:11.39(Nuclear Science & Technology)In the design of innovative nuclear reactors such as fast reactors, the improvement of the prediction accuracies for neutronics properties is an important task. The nuclear data adjustment is a promising methodology for this issue. The idea of the nuclear data adjustment was first proposed in 1964. Toward its practical application, however, a great deal of study has been conducted over a long time. While it took about 10 years to establish the theoretical formulation, the research and development for its practical application has been conducted for more than half a century. Researches in this field are still active, and the fact suggests that the improvement of the prediction accuracies is indispensable for the development of new types of nuclear reactors. Massimo Salvatores, who passed away in March 2020, was one of the first proposers to develop the nuclear data adjustment technique, as well as one of the great contributors to its practical application. Reviewing his long-time works in this area is almost the same as reviewing the history of the nuclear data adjustment methodology. The authors intend that this review would suggest what should be done in the future toward the next development in this area. The present review consists of two parts: a) the establishment of the nuclear data adjustment methodology and b) the achievements related to practical applications. Furthermore, the former is divided into two aspects: the study on the nuclear data adjustment theory and the numerical solution for sensitivity coefficient that is requisite for the nuclear data adjustment. The latter is separated to three categories: the use of integral experimental data, the uncertainty quantification and design target accuracy evaluation, and the promotion of nuclear data covariance development.
Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Otsuka, Noriaki; Tsuchiya, Kunihiko
Journal of Nuclear Science and Technology, 57(12), p.1276 - 1286, 2020/12
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)The influence of materials of mineral-insulated (MI) cables on their electrical characteristics upon exposure to high-temperature conditions was examined via a transmission test, in the objective of achieving the stability of the potential distribution along the cable length. Occurrence of a voltage drop along the cable was confirmed for aluminum oxide (AlO
) and magnesium oxide (MgO), as insulating materials of the MI cable. A finite-element method (FEM)-based analysis was performed to evaluate the leakage in the potentials, which was found at the terminal end. Voltage drop yields by the transmission test and the analysis were in good agreement for the MI cable of Al
O
and MgO materials, which suggests the reproducibility of the magnitude relationship of the experimental results via the FEM analysis. To suppress the voltage drop, the same FEM analysis was conducted, the diameter of the core wires (
) and the distance between them (
) were varied. Considering the variation of
, the potential distribution in the MI cable produced a minimum voltage drop corresponding to a ratio
of 0.35, obtained by dividing
with that of the insulating material (
). In case of varying
, a minimum voltage drop was l/
of 0.5.
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.
Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio
Nuclear Engineering and Design, 233(1-3), p.89 - 101, 2004/10
Times Cited Count:10 Percentile:55.11(Nuclear Science & Technology)no abstracts in English
Shiroya, Seiji*; Misawa, Tsuyoshi*; Unesaki, Hironobu*; Ichihara, Chihiro*; Kobayashi, Keiji*; Nakamura, Hiroshi*; Shin, Kazuo*; Imanishi, Nobutsugu*; Kanazawa, Satoshi*; Mori, Takamasa
JAERI-Tech 2004-025, 93 Pages, 2004/03
In view of the future plan of Research Reactor Institute, Kyoto University, the present study consisted of (1) the transmission experiments of high energy neutrons through materials, (2) experimental simulation of ADSR using the Kyoto University Critical Assembly(KUCA), and (3) conceptual neutronics design study on KUR type ADSR using the MCNP-X code. Through the present study, valuable knowledge on the basic nuclear characteristics of ADSR, which is indispensable to promote the study on ADSR, was obtained both theoretically and experimentally. For the realization of ADSR, it is considered to be necessary to accumulate results of research steadily. For this purpose, it is inevitable (1) to compile the more precise nuclear data for the wide energy range, (2) to establish experimental techniques for reactor physics study on ADSR including subcriticality measurement and absolute neutron flux measurement, and (3) to develop neutronics calculation tools which take into account the neutron generation process by the spallation reaction and the delayed neutron behavior.
Kugo, Teruhiko; Tsuchihashi, Keiichiro*; Nakakawa, Masayuki; Ido, Masaru*
JAERI-Data/Code 2000-011, p.138 - 0, 2000/02
no abstracts in English
Yamamoto, Nobuo
JAERI-Data/Code 99-023, 65 Pages, 1999/04
no abstracts in English
Kugo, Teruhiko; Nakagawa, Masayuki;
Journal of Nuclear Science and Technology, 34(8), p.760 - 770, 1997/08
Times Cited Count:2 Percentile:22.91(Nuclear Science & Technology)no abstracts in English
Kugo, Teruhiko; Nakagawa, Masayuki
PHYSOR 96: Int. Conf. on the Physics of Reactors, 1, p.B73 - B81, 1996/00
no abstracts in English
Yamamoto, Nobuo;
JAERI-Data/Code 95-018, 634 Pages, 1995/12
no abstracts in English
Kugo, Teruhiko; Nakagawa, Masayuki
Transactions of the American Nuclear Society, 73, p.207 - 208, 1995/00
no abstracts in English
Research Committee on Reactor Physics
JAERI-M 93-254, 36 Pages, 1994/01
no abstracts in English
I.Smid*; Akiba, Masato; Araki, Masanori; ; Sato, Kazuyoshi
JAERI-M 93-149, 200 Pages, 1993/07
no abstracts in English
Research Committee on Reactor Physics
JAERI-M 92-209, 43 Pages, 1993/01
no abstracts in English
; Nakagawa, Masayuki; Mori, Takamasa; Kugo, Teruhiko
Proc. of the Int. Conf. on Design and Safety of Advanced Nuclear Power Plants,Vol. 3, p.32.1_1 - 32.1_8, 1992/00
no abstracts in English
Ishihara, Masahiro; Iyoku, Tatsuo; ; ; Shiozawa, Shusaku
JAERI-M 91-154, 39 Pages, 1991/10
no abstracts in English
Ishihara, Masahiro; Iyoku, Tatsuo; ; ; Shiozawa, Shusaku
JAERI-M 91-153, 51 Pages, 1991/10
no abstracts in English
JAERI-M 85-035, 409 Pages, 1985/03
no abstracts in English
; Ichikawa, Michio
JAERI-M 84-113, 35 Pages, 1984/06
no abstracts in English