Wang, H.*; Yu, H.*; Kondo, Sosuke*; Okubo, Nariaki; Kasada, Ryuta*
Corrosion Science, 175, p.108864_1 - 108864_12, 2020/10
Corrosion tests were performed on newly developed alumina-forming austenitic (AFA) steels in stagnant lead bismuth eutectic (LBE) with saturated and low oxygen concentrations at 450C for 430 h. The steels exhibited enhanced corrosion resistance to the LBE environments with the increasing of Al content. A continuous and protective Al-rich oxide scale formed on the steel specimens that were exposed to LBE with a low oxygen concentration, whereas a non-protective and stratified oxide scale formed in the oxygen saturated LBE.
Kanai, Akihiko*; Kasada, Ryuta*; Nakajima, Motoki; Hirose, Takanori; Tanigawa, Hisashi; Enoeda, Mikio; Konishi, Satoshi*
Journal of Nuclear Materials, 455(1-3), p.431 - 435, 2014/12
Kanai, Akihiko*; Park, C.*; Noborio, Kazuyuki*; Kasada, Ryuta*; Konishi, Satoshi*; Hirose, Takanori; Nozawa, Takashi; Tanigawa, Hiroyasu
Fusion Engineering and Design, 89(7-8), p.1653 - 1657, 2014/10
Kim, B. J.; Kasada, Ryuta*; Kimura, Akihiko*; Wakai, Eiichi; Tanigawa, Hiroyasu
Journal of Nuclear Materials, 442(1-3), p.S38 - S42, 2013/11
Tokunaga, Tomonori*; Watanabe, Hideo*; Yoshida, Naoaki*; Nagasaka, Takuya*; Kasada, Ryuta*; Lee, Y.-J.*; Kimura, Akihiko*; Tokitani, Masayuki*; Mitsuhara, Masatoshi*; Hinoki, Tatsuya*; et al.
Journal of Nuclear Materials, 442(1-3), p.S287 - S291, 2013/11
Kwon, S.*; Sato, Satoshi; Kasada, Ryuta*; Konishi, Satoshi*
Fusion Science and Technology, 64(3), p.599 - 603, 2013/09
Tritium production/breeding behavior in LiPb blanket module was evaluated by neutron transport code MCNP with nuclear cross-section data from FENDL-2.1 libraries. The calculation results were suggested that the sufficient TBR can be obtained in the SiC-LiPb blanket concept. A proper integral experiment on LiPb with DT neutrons in a small test module was evaluated. Also, tritium breeding ratio, tritium production ratio, proper neutron shielding material and nuclear heating in the module were evaluated. With the results of TPR and actual neutron generation devices, we have proposed the plan of the integral experiment and measurable tritium amount.
Kasada, Ryuta*; Goto, Takuya*; Fujioka, Shinsuke*; Hiwatari, Ryoji*; Oyama, Naoyuki; Tanigawa, Hiroyasu; Miyazawa, Junichi*; Young Scientists Special Interest Group on Fusion Reactor Realization*
Purazuma, Kaku Yugo Gakkai-Shi, 89(4), p.193 - 198, 2013/04
Japanese young researchers who have interest in realizing fusion reactor have analyzed Technology Readiness Levels (TRL) in Young Scientists Special Interest Group on Fusion Reactor Realization. In this report, brief introduction to TRL assessment and a view of TRL assessment against fusion reactor projects conducting in Japan.
Wakai, Eiichi; Kim, B. J.; Nozawa, Takashi; Kikuchi, Takayuki; Hirano, Michiko*; Kimura, Akihiko*; Kasada, Ryuta*; Yokomine, Takehiko*; Yoshida, Takahide*; Nogami, Shuhei*; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 6 Pages, 2013/03
Wakai, Eiichi; Nogami, Shuhei*; Kasada, Ryuta*; Ito, Yuzuru*; Takada, Fumiki; 6 of others*
Journal of Nuclear Materials, 417, p.1325 - 1330, 2011/10
Kimura, Akihiko*; Kasada, Ryuta*; Iwata, Noriyuki*; Kishimoto, Hirotatsu*; Zhang, C. H.*; Isselin, J.*; Dou, P.*; Lee, J. H.*; Muthukumar, N.*; Okuda, Takanari*; et al.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9220_1 - 9220_8, 2009/05
Cladding material development is essential for realization of highly efficient high burn-up operation of next generation nuclear systems, where high performance is required for the materials, that is, high strength at elevated temperature, high resistance to corrosion and high resistance to irradiation. Oxide dispersion strengthening (ODS) ferritic steels are considered to be most adequate for the cladding material because of their high strength at elevated temperature. In this work, "Super ODS steel" that has better corrosion resistance than 9Cr-ODS steel, has been developed for application to cladding of a variety of next generation nuclear systems. In the following ten papers, the recent experimental results of "Super ODS steel" R&D will be presented, indicating that many unexpected preferable features were found in the mechanical properties of nano-sized oxide dispersion high-Cr ODS ferritic steel. A series of paper begins with alloy design of "Super ODS steel". Corrosion issue requires Cr concentration more than 14wt.%, but aging embrittlement issue requires less than 16wt.%. An addition of 4wt.%Al is effective to improve corrosion resistance of 16wt.%Cr-ODS steel in supercritical water (SCW) and lead-bismuth eutectic (LBE), while it is detrimental to high-temperature strength. Additions of 2wt.%W and 0.1wt.%Ti are necessary to keep high strength at elevated temperatures. An addition of small amount of Zr or Hf results in a significant increase in creep strength at 700 C in Al added ODS steels. Tube manufacturing was successfully done for the super ODS steel candidates. "Super ODS steel" is promising for the fuel cladding material of next generation nuclear systems, and the R&D is now ready to proceed to the next stage of empirical verification.
Lee, J. H.*; Kimura, Akihiko*; Kasada, Ryuta*; Iwata, Noriyuki*; Kishimoto, Hirotatsu*; Zhang, C. H.*; Isselin, J.*; Dou, P.*; Muthukumar, N.*; Okuda, Takanari*; et al.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9223_1 - 9223_6, 2009/05
Corrosion is a critical issue for cladding materials, especially, in sever corrosion environment as supercritical pressurized water (SCPW). In this work, the effects of alloy elements on the corrosion behavior in SCPW were investigated for a series of oxide dispersion strengthened (ODS) steels to design alloy compositions for corrosion resistant super ODS ferritic steels. Corrosion tests were carried out for the ODS steels with different concentrations of Cr and Al in SCPW at 773 K at 25 MPa with 8 ppm of dissolved oxygen. The corrosion rate of SUS430, which contained 16wt.%Cr, was much higher than 16Cr-ODS steel. This suggests that nano-sized oxide particles dispersion and very fine grains play an important role in suppression of the corrosion. The corrosion of the ODS steel was reduced by an addition of Al in 16wt.%Cr-ODS steel but not in 19Cr-ODS steel. FE-EPMA chemical analysis clearly indicated that the surface of the Al added ODS steels was covered by alumina which suppresses the corrosion in SCPW. It is considered that an adequate combination of the contents of Cr and Al is ranging (14-16)Cr and (3.5-4.5)Al.
Kasada, Ryuta*; Lee, S. G.*; Lee, J. H.*; Omura, Takamasa*; Zhang, C. H.*; Dou, P.*; Isselin, J.*; Kimura, Akihiko*; Inoue, Masaki; Ukai, Shigeharu*; et al.
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9072_1 - 9072_5, 2009/05
The newly-developed Al-added ODS ferritic steels with an addition of Zr or Hf, socalled super ODS candidate steels, showed good notch-impact properties in the as-received condition with keeping the excellent creep strength.
Kishimoto, Hirotatsu*; Kasada, Ryuta*; Kimura, Akihiko*; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*
Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9219_1 - 9219_8, 2009/05
The Super ODS steels, having excellent high-temperature strength and highly corrosion resistant, are considered to increase the energy efficiency by higher temperature operation and extend the lifetime of next generation nuclear systems. High-temperature strength of the ODS steels strongly depends on the dispersion of oxide particles, therefore, the irradiation effect on the dispersed oxides is critical in the material development. In the present research, ion irradiation experiments were employed to investigate microstructural stability under the irradiation environment at elevated temperatures. Ion irradiation experiments were performed with 6.4 MeV Fe ions irradiated at 650 C up to a nominal displacement damage of 60 dpa. Microstructural investigation was carried out using TEM and EDX. No significant change of grains and grain boundaries was observed by TEM investigation after the ion irradiation. Main oxide particles in the 16Cr-4Al-0.1Ti (SOC-1) ODS steel were (Y, Al) complex oxides. (Y, Ti) complex oxides were in 16Cr-0.1Ti (SOC-5) and 15.5Cr-2W-0.1Ti (SOCP-3). (Y, Zr) complex oxides were in 15.5Cr-4Al-0.6Zr (SOCP-1). No significant modification of these complex oxides was detected after the ion irradiation up to 60 dpa at 650 C. The stable complex oxides are considered to keep highly microstructural stability of the Super ODS steels under the irradiation environments.
Nishitani, Takeo; Tanigawa, Hiroyasu; Jitsukawa, Shiro; Nozawa, Takashi; Hayashi, Kimio; Yamanishi, Toshihiko; Tsuchiya, Kunihiko; Mslang, A.*; Baluc, N.*; Pizzuto, A.*; et al.
Journal of Nuclear Materials, 386-388, p.405 - 410, 2009/04
The establishment of the breeding blanket technology is one of the most important engineering issues on the DEMO development. For the DEMO blanket, developments of the structural materials and functional materials such as tritium breeder and neutron multiplier. Which should be used under the savior circumstance such as high neutron fluence, high temperature and strong magnetic field, are urgent issues. In the Broader Approach activities initiated by EU and Japan, developments of reduced activation ferritic martensitic steels as a DEMO blanket structural material, SiC/SiC composites, advanced tritium breeders and neutron multiplier for DEMO blankets, are planed as common interest issues of EU and Japan. This paper describes the overview of the development program.
Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki; Kasada, Ryuta*; Wakai, Eiichi; Serizawa, Hisashi*; Kawahito, Yosuke*; Jitsukawa, Shiro; Kimura, Akihiko*; Kono, Yutaka*; et al.
Fusion Engineering and Design, 83(10-12), p.1471 - 1476, 2008/12
Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. F82H, which were developed and studied in Japan, was designed with an emphasis on high temperature properties and weldability. The database on F82H properties is currently the most extensive available among the existing RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER Test Blanket Module (TBM) suggested by recent achievements in Japan.
Hayashi, Takao; Kasada, Ryuta*; Tobita, Kenji; Nishio, Satoshi; Sawai, Tomotsugu; Tanigawa, Hiroyasu; Jitsukawa, Shiro
Fusion Engineering and Design, 82(15-24), p.2850 - 2855, 2007/10
The impact of increasing the enrichment of N in low activation ferritic steel, F82H, of a fusion reactor has been investigated in order to increase the fraction of low level material (LLM), which can be disposed by shallow land burial. Carbon-14, mainly produced from nitrogen, is one of the most critical nuclei for qualifying as a LLM. The concentration of nitrogen in F82H is 200 ppm in the calculations. The enrichment of N was varied from natural abundance of 0.37% to 95%. The concentration of C at the outboard first wall decreased from 7.8 10 to 3.2 10 Bq/g by enriching N, which is lower than the C regulation (3.7 10 Bq/g) for LLM in Japan. In the permanent blanket, the highest C concentration with 95% N enriched nitrogen was 1.0 10 Bq/g. The C concentration on the inboard side was lower than the outboard side. Therefore, with regard to the C concentration, the F82H used in the inboard and outboard blankets can qualify as a LLM by enriching N.
Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Hirose, Takanori; Kasada, Ryuta*; Wakai, Eiichi; Jitsukawa, Shiro; Kimura, Akihiko*; Koyama, Akira*
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 6 Pages, 2007/03
The status of research and development of reduced activation martensitic steels (RAMs) in Japan are reviewed and key issues suggested from recent achievements in Japan since the last conference are highlighted, with the aim of the fabrication for the ITER Test Blanket Module (TBM) and application for the DEMO reactor. It was pointed out that international collaboration would be desirable for research on key issues such as precipitate stability under irradiation or Ta effects which are common for all RAMs and require an extensive research effort.
Kimura, Akihiko*; Kasada, Ryuta*; Kishimoto, Hirotatsu*; Okuda, Takanari*; Inoue, Masaki; Abe, Fujio*; Onuki, Somei*; Ukai, Shigeharu*; Fujisawa, Toshiharu*
no journal, ,
The project named as "Super ODS Steel Research and Development towards Highly Efficient Nuclear Systems" will be reviewed. Also, experimental results of mechanical alloying process will be topically introduced.
Iwata, Noriyuki*; Muthukumar, N.*; Kasada, Ryuta*; Kimura, Akihiko*; Okuda, Takanari*; Inoue, Masaki; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*
no journal, ,
Corrosion behavior of Super ODS steels in SCPW in the project named as "Super ODS Steel Research and Development towards Highly Efficient Nuclear Systems" will be summarized.
Kishimoto, Hirotatsu*; Kasada, Ryuta*; Kimura, Akihiko*; Onuki, Somei*; Okuda, Takanari*; Abe, Fujio*; Inoue, Masaki; Fujisawa, Toshiharu*
no journal, ,
Irradiation effects on Super ODS steels will be summarized for the project named as "Super ODS Steel Research and Development towards Highly Efficient Nuclear Systems".