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Ebihara, Kenichi; Sekine, Daiki*; Sakiyama, Yuji*; Takahashi, Jun*; Takai, Kenichi*; Omura, Tomohiko*
International Journal of Hydrogen Energy, 48(79), p.30949 - 30962, 2023/09
Times Cited Count：0 Percentile：0.01(Chemistry, Physical)To understand hydrogen embrittlement (HE), which is one of the stress corrosion cracking of steel materials, it is necessary to know the H distribution in steel, which can be effectively interpreted by numerical simulation of thermal desorption spectra. In weld metals and TRIP steels, residual austenite significantly influences the spectra, but a clear H distribution is not well known. In this study, an originally coded two-dimensional model was used to numerically simulate the previously reported spectra of high-carbon ferritic-austenitic duplex stainless steels, and it was found that H is mainly trapped at the carbide surface when the amount of H in the steel is low and at the duplex interface when the amount of H is high. It was also found that the thickness dependence of the H desorption peak for the interface trap site is caused by a different reason than the conventional one.
Kinoshita, Hidetaka; Kaminaga, Masanori; Haga, Katsuhiro; Hino, Ryutaro
JAERI-Tech 2002-052, 28 Pages, 2002/06
JAERI-Tech-2002-052.pdf:7.59MB
Since the Neutron Scattering Facility will be using mercury as the target material and contain radioactive products, it is necessary to estimate reliability of instruments in a system. The system would be damaged by erosion. An erosion test section and coupons were installed in the mercury loop, and their thickness was measured. As a result, the erosion is about 3
m in 1000 hours under 0.7m/s condition. The wall thickness decrease during facility lifetime of 30 years is estimated to be less than 0.5mm. Therfore, the effect of erosion on component strength is extremely small. Moreover, a measurement of residual mercury on the piping surface was carried out. As a result, 19g/m
was obtained. Thus, estimation of residual mercury for 150A-sch80 piping is 8.5g/m, and for the mercury target is about 40g. As for the target, radioactivity of the residual mercury is 1.2
10
Bq, which is extremely lower than that in the target casing of 1.010
Bq. Then, there is no influence for maintenance and storage of the spent mercury target.
Aoyagi, Takayoshi*; *; Mihara, Morihiro; Okutsu, Kazuo*; Maeda, Munehiro*
JNC TN8400 2001-024, 103 Pages, 2001/06
JNC-TN8400-2001-024.pdf:8.84MB
In the disposal concept of TRU waste, concentrated disposal of wastes forms in large cross-section underground cavities is envisaged, because most of TRU waste is no-heat producing in spite of large generated volume as compared with HLW. In the design of engineered barrier system based on large cross-section cavities, it is necessary to consider the long-term mechanical process such as creep displacement of the host rock from the viewpoint of the stability of engineered barrier system. In this study, the long-term creep displacement of the host rock was calculated using the non-linear viscoelasticity model and the effects on the stability of engineered barrier system was evaluated. As a result, in the disposal concept of crystalline rock, no creep displacement occurred at the time after 1 milion year. On the other hand, in the disposal concept of sedimentary rock, creep displacement of 80
90mm occurred at the time after 1 milion year. Also, in this calculation, a maximum reduction of 45mm concerned with the thickness of buffer material was estimated. But these values resulted within allowance of design values. Therefore, these results show that the effects of the creep displacement on the stability of engieered barrier system would not be significant.
Hidaka, Akihide; Shibazaki, Hiroaki*; Yoshino, T.*; Sugimoto, Jun
Journal of Aerosol Science, 31(9), p.1045 - 1059, 2000/09
Times Cited Count：1 Percentile：18.21(Engineering, Chemical)no abstracts in English
Hidaka, Akihide; Shibazaki, Hiroaki*; Maruyama, Yu; Yoshino, T.*; Sugimoto, Jun
NEA/CSNI/R(98)4, 14 Pages, 2000/02
no abstracts in English
Shibamoto, Yasuteru; Kukita, Yutaka*; Nakamura, Hideo; Park, H. S.*; Anoda, Yoshinari
Proceedings of 8th International Conference on Nuclear Engineering (ICONE-8) (CD-ROM), p.12 - 0, 2000/00
no abstracts in English
; ; Ishiguro, Katsuhiko; Nakajima, Kunihiko*;
JNC TN8400 99-087, 41 Pages, 1999/11
Corrosion of the carbon steel overpack leads to a volume expansion since the specific gravity of corrosion products is smaller than carbon steel. The buffer material is compressed due to the corrosive swelling, reducing its thickness and porosity. On the other hand, Buffer material may be extruded into fractures of the surrounding rock and this may lead to a deterioration of the planned functions of the buffer, including retardation of nuclides migration and colloid filtration. In this study, the sensitivity analyses for the effect of volume expansion and intrusion of the buffer material on nuclide migration in the engineering barrier system are carried out. The sensitivity analyses were performed on the decrease in the thickness of the buffer material in the radial direction caused by the corrosive swelling, and the change in the porosity and dry density of the buffer caused by both compaction due to corrosive swelling and intrusion of buffer material. As results, it was found the maximum release rates of relatively shorter half-life nuclides from the outside of the buffer material decreased for taking into account of a volume expansion due to overpack corrosion. On the other hand, the maximum release rates increased when the intrusion of buffer material was also taking into account. It was, however, the maximum release rates of longer half-life nuclides, such as Cs-137 and Np-237, were insensitive to the change of buffer material thickness, and porosity and dry density of buffer.
*; *; Tanai, Kenji
JNC TN8400 99-047, 54 Pages, 1999/11
This paper reports on the design process for a carbon-steel overpack as a key component in the engineered barrier system of a deep geological repository described in the 2nd progress report. The results of the research and development regarding design requirements, configuration, manufacturing and inspection of overpack are also described. The concept of a composite overpack composed of two different materials is also considered. First, the design requirements for an overpack and presume environmental and design conditions for a repository are provided. For a candidate material of carbon steel overpack, forging material is selected considering enough experience of using this material in nuclear power boilers and other components. Second, loading conditions after emplacement in a repository are set and the pressure-resistant thickness of overpack is calculated. The corrosion thickness to achieve an assigned 1000 year life time and the required thickness to prevent radiolysis of ground water which might enhance corrosion rate are also determined. As aresult, the total required thickness of a carbon-steel overpack is conservatively estimated to 190 mm. This is a reduction of about 30% from the previous estimate provided in the 1st Progress Report. Additional items that must be considered in manufacturring and operating overpacks (i.e. sealing of vitrified waste, examination of main body and sealing welding, mechanism of handling) are evaluated on the basis of current technology, specific future data needs are identified. With respect to the concept of composite overpack (i.e., an outer vessel to provide corrosion-allowance or corrosion-resistant performance and an inner vessel to provide pressure-resistance), the differences in design concepts between the carbon-steel overpack and such composite overpacks are analyzed. Future data needs and analytical capabilities with respect to overpacks are also summarized.
Sato, Masayasu; Isei, Nobuaki; Ishida, Shinichi; Isayama, Akihiko
Journal of the Physical Society of Japan, 67(9), p.3090 - 3099, 1998/09
Times Cited Count：12 Percentile：62.22(Physics, Multidisciplinary)no abstracts in English
; Ara, Katsuyuki
Denki Gakkai Magunethikkusu Kenkyukai Shiryo, p.37 - 41, 1997/07
no abstracts in English
Sato, Satoshi; Takatsu, Hideyuki; *; *; Iida, Hiromasa; R.Santoro*
Fusion Technology 1996, 0, p.1587 - 1590, 1997/00
no abstracts in English
*; *; Matsubayashi, Masahito; Mori, Chizuo*
Nihon Genshiryoku Gakkai-Shi, 39(8), p.647 - 656, 1997/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Sato, Masayasu; Isei, Nobuaki; Isayama, Akihiko; Ishida, Shinichi; Shirai, Hiroshi
Proc. of 1996 Int. Conf. on Plasma Physics, Vol.2, p.1438 - 1441, 1997/00
no abstracts in English
; Ara, Katsuyuki
MAG-96-265, 0, p.11 - 24, 1996/12
no abstracts in English
Ishikawa, Isamu
Proc. of 11th KAIF/KNS Annual Conf., 0, p.611 - 619, 1996/00
no abstracts in English
Ara, Katsuyuki; Sakasai, Kaoru; Kishimoto, Maki; ;
Nihon Oyo Jiki Gakkai-Shi, 19, p.493 - 496, 1995/00
no abstracts in English
Ara, Katsuyuki; Sakasai, Kaoru; Kishimoto, Maki; ;
MAG-94-24, p.43 - 52, 1994/03
no abstracts in English
Ara, Katsuyuki; ; ; Sakasai, Kaoru
Proc. of the Japan-Central Europe Joint Workshop on Advanced Computing in Engineering, 0, p.221 - 226, 1994/00
no abstracts in English
Genka, Tsuguo
Isotope News, 0(470), p.12 - 15, 1993/08
no abstracts in English
Genka, Tsuguo
Radioisotopes, 42(7), p.437 - 438, 1993/00
no abstracts in English
