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Mitsuhara, Masatoshi*; Kurino, Koichi*; Yano, Yasuhide; Otsuka, Satoshi; Toyama, Takeshi*; Onuma, Masato*; Nakashima, Hideharu*
Tetsu To Hagane, 109(3), p.189 - 200, 2023/03
Times Cited Count：0 Percentile：0(Metallurgy & Metallurgical Engineering)Oxide Dispersion Strengthened (ODS) ferritic steel, a candidate material for fast reactor fuel cladding, has low thermal expansion, good thermal conductivity, and excellent resistance to irradiation damage and high temperature strength. The origin of the excellent high-temperature strength lies in the dispersion of fine oxides. In this study, creep tests at 700 or 750
C, which are close to the operating temperatures of fast reactors, and high-temperature tensile tests at 900 to 1350 C, which simulate accident conditions, were conducted on 9Cr ODS ferritic steels, M11 and MP23, and 12Cr ODS ferritic steel, F14, to confirm the growth behavior of oxides. In the M11 and F14 creep test samples, there was little oxide growth or decrease in number density from the initial state, indicating that dispersion strengthening by oxides was effective during deformation. After creep deformation of F14, the development of dislocation substructures such as dislocation walls and subgrain boundaries was hardly observed, and mobile dislocations were homogeneously distributed in the grains. The dislocation density increased with increasing stress during the creep test. In the high-temperature ring tensile tests of MP23 and F14, the strength of both steels decreased at higher temperatures. In MP23, elongation decreased with increasing test temperature from 900 to 1100 C, but increased at 1200 C, decreased drastically at 1250 C, and increased again at 1300 C. In F14, elongation decreased with increasing temperature. It was inferred that the formation of the 
-ferrite phase was responsible for this complex change in mechanical properties of MP23 from 1200 to 1300 C.
Oka, Hiroshi*; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Hashimoto, Naoyuki*
Journal of Nuclear Materials, 572, p.154032_1 - 154032_8, 2022/12
Times Cited Count：3 Percentile：66.21(Materials Science, Multidisciplinary)9Cr oxide dispersion strengthened steels with slightly different nitrogen concentrations (0.0034 - 0.029 wt%) were prepared and their creep property at 973 K was investigated with microstructural characterization before and after the creep test. The creep strength decreased significantly as the nitrogen concentration increased. Microstructural observation revealed that, in the higher nitrogen concentration specimen, coarse Y-rich inclusions were found along the boundary between transformed ferrite region and residual ferrite region. The solubility difference of nitrogen in 
and 
phase would induce the localized increment of nitrogen concentration in the boundary region during the austenitizing process, resulting in the thermodynamic destabilization and subsequent coarsening of the dispersed oxide particles. The rows of creep voids were found near the rupture part of the crept specimen, suggesting that the coarse inclusions were the starting point of creep void formation and the subsequent premature fracture.
Yano, Yasuhide; Hashidate, Ryuta; Tanno, Takashi; Imagawa, Yuya; Kato, Shoichi; Onizawa, Takashi; Ito, Chikara; Uwaba, Tomoyuki; Otsuka, Satoshi; Kaito, Takeji
JAEA-Data/Code 2021-015, 64 Pages, 2022/01
JAEA-Data-Code-2021-015.pdf:2.6MB
From a view point of practical application of fast breeder reactor cycles, which takes advantage of safety and economic efficiency and makes a contribution of volume reduction and mitigation of degree of harmfulness of high-level radioactive waste, it is necessary to develop fuel cladding materials for fast reactors (FRs) in order to achieve high-burnup. Oxide dispersion strengthened (ODS) steel have been studied for use as potential fuel cladding materials in FRs owing to their excellent resistance to swelling and their high-temperature strength in Japan Atomic Energy Agency. It is very important to establish the materials strength standard in order to apply ODS steels as a fuel cladding. Therefore, it is necessary to acquire the mechanical properties such as tensile, creep rupture strength tests and so on. In this study, tensile and creep rupture strengths of 9Cr-ODS steel claddings were evaluated using by acquired these data. Because of the phase transformation temperature of 9Cr-ODS steel, temperature range for the evaluation was divided into two ones at AC1 transformation temperature of 850C.
Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Kaito, Takeji; Ukai, Shigeharu*
Materials Transactions, 62(8), p.1239 - 1246, 2021/08
Times Cited Count：5 Percentile：39.91(Materials Science, Multidisciplinary)The FeCrAl-ODS alloy claddings were manufactured and Vickers hardness, ring tensile tests and TEM observations of these claddings were performed to investigate the effects of thermal aging at 450 C for 5,000 and 15,000 h. The age-hardening of all FeCrAl-ODS alloy cladding was found. In addition, the significant increase in tensile strength was accompanied by much larger loss of ductility. It was suggested that this age-hardening behavior was attributed to the (Ti, Al)-enriched phase (
' phase) and the ' phase precipitates (content of Al is 
7 wt%). In comparison with FeCrAl-ODS alloys with almost same chemical compositions, there was significant age-hardening in both alloys. However, the extrusion bar with no-recrystallized structures was keeping good ductility. It was suggested that this different behavior of reduction ductility was attributed to the effects of grain boundaries, dislocation densities and specimen preparation direction.
Ukai, Shigeharu*; Yano, Yasuhide; Inoue, Toshihiko; Sowa, Takashi*
Materials Science & Engineering A, 812, p.141076_1 - 141076_11, 2021/04
Times Cited Count：12 Percentile：71.72(Nanoscience & Nanotechnology)FeCrAl oxide dispersion strengthened alloys are promising materials for accident tolerant fuels for light water reactors (LWRs). In these alloys, Al and Cr are key elements with important synergistic effects: enhancement of the formation of oxidation-resistant Al
O
phase by Cr addition and suppression of the formation of the embrittling Cr-rich ' phase by Al addition. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The Al and Cr contents were systematically varied from 9-16 at.% and 10-17 at.%, respectively, and tensile tests were conducted at 298 K, 573 K and 973 K in the as-annealed condition. The solid solution strengthening increased linearly, 20 MPa per 1 at.% Al and 5 MPa per 1 at.% Cr, at the typical LWR operational temperature of 573 K. The conventional Fleischer-Friedel and Labusch theories cannot explain this level of solid-solution strengthening. It was shown that Suzuki's double kink theory for screw dislocations reasonably predicts the solid solution strengthening by Al and Cr as well as the inverse dependency on the absolute temperature and linear dependency on the Al and Cr content.
Tanno, Takashi; Takeuchi, Masayuki; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Materials, 494, p.219 - 226, 2017/10
Times Cited Count：17 Percentile：85.05(Materials Science, Multidisciplinary)Oxide dispersion strengthened (ODS) steel cladding tubes have been developed for fast reactors. 9 chromium ODS and 11Cr-ODS tempered martensitic steels are prioritized for the candidate material in research being carried out at JAEA. In this work, fundamental immersion tests and electro-chemical tests of 9 to 12Cr-ODS steels were systematically conducted in various nitric acid solutions at 95C. The corrosion rate exponentially decreased with effective solute chromium concentration (Cr
) and nitric acid concentration. Addition of oxidizing ions also suppressed the corrosion rate. According to polarization curves and surface observations in this work, the combination of low Cr and dilute nitric acid could not prevent the active dissolution at the beginning of immersion, and the corrosion rate was high. In comparison, higher Cr, concentrated nitric acid and addition of oxidizing ions helped to prevent the active dissolution, and suppressed the corrosion rate.
Tanno, Takashi; Yano, Yasuhide; Oka, Hiroshi; Otsuka, Satoshi; Uwaba, Tomoyuki; Kaito, Takeji
Nuclear Materials and Energy (Internet), 9, p.353 - 359, 2016/12
Times Cited Count：9 Percentile：64.39(Nuclear Science & Technology)Materials for core components of fusion reactors and fast reactors, such as blankets and fuel cladding tubes, must be excellent in high temperature strength and irradiation resistance because they will be exposed to high heat flux and heavy neutron irradiation. Oxide dispersion strengthened (ODS) steels have been developing as the candidate material. Japan Atomic Energy Agency (JAEA) have been developing 9 and 11 Chromium (Cr) ODS steels for advanced fast reactor cladding tubes. The JAEA 11Cr-ODS steels were rolled in order to evaluate their anisotropy. Tensile tests and creep tests of them were carried out at 700 C in longitudinal and transverse orientation. The anisotropy of tensile strength was negligible, though that of creep strength was distinct. The observation results and chemical composition analysis suggested that the cause of the anisotropy in creep strength was prior powder boundary including Ti-rich precipitates.
Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Onuma, Masato*
no journal, ,
Microstructural changes of 9Cr-ODS steel introduced by thermo-mechanical treatment (hot isostatic press (HIP), hot extrusion, and hot forge) were evaluated. The weight-ratio of residual ferrite phase reduced and the number density of nano-oxide particle decreased after the thermo-mechanical treatment. The change of number density and size of nano-oxide particle would affect the weight-ratio of residual ferrite phase because the formation of the residual ferrite phase is led by the pin effect of the nano-oxide particle.
Inoue, Toshihiko; Sekio, Yoshihiro; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Furukawa, Tomohiro; Kaito, Takeji; Torimaru, Tadahiko*; Hayashi, Shigenari*; et al.
no journal, ,
In order to evaluate the strength and deformation in severe accident, the transient burst tests were carried out with various heating rates (from 0.1 to 10 K/s) and hoop stresses (from 50 to 200 MPa) to provide more evaluation data. The test materials were core materials in fast reactors, 9-18Cr-ODS and accident tolerant fuel cladding tube in the light water reactors, FeCrAl-added ODS ferritic steels. Result, the rupture strength dropped with increasing hoop stress and decreasing heating rate. The burst strength of Al-added ODS steels was lower than other ODS steels, Al and Zr-added ODS steels show good transient burst strength.
Maeda, Seiichiro
no journal, ,
The uranium-plutonium mixed oxide (MOX) fuel bearing several percent minor actinides is adopted as the fast reactor fuel to reduce the high level radioactive waste volume and its toxicities. Their physical properties such as melting points and thermal conductivities are investigated symmetrically. It is getting clear that the effect of MA bearing is not so significant. The MOX fuel pins including up to 5% of Am were irradiated in Joyo to confirm their irradiation performance. Some types of irradiation experiments using MA-MOX fuels are planned after Joyo restart. The obtained data will contribute to develop fuel pin performance codes for MA-MOX fuels. SmART (Small Amount of Reuse Fuel Test) cycle program is in progress to demonstrate MA recycles starting from FR spent fuels. Furthermore, long life cladding with ODS steel is under development to enhance the transmutation efficiency of MA in one cycle. Thus, R&Ds in FR systems advances steadily to solve the waste issue.
Yuzawa, Sho*; Yabuuchi, Kiyohiro*; Kimura, Akihiko*; Sakamoto, Kan*; Hirai, Mutsumi*; Yamashita, Shinichiro
no journal, ,
Oxide dispersion strengthened steel with high Cr and Al concentration in chemical composition (FeCrAl-ODS steel) has been proposed as a promising candidate for the accident tolerant fuel cladding of light water reactors (LWRs) because of their excellent oxidation and corrosion resistance under high temperature water and steam environments. Neither there are no sufficient knowledge on welding technology of FeCrAl-ODS nor it is known that aluminum addition remarkably degrades the weldability of the ODS steel. In this study, electron beam (EB) welding and tungsten inert gas (TIG) welding were applied to the SUS430 endcap welding to cladding tube made of FeCrAl-ODS steel and performed the bonding strength and corresponding damage structure evaluations at the bonding part.
Suzuki, Akihiro*; Onuma, Masato*; Tanno, Takashi; Oka, Hiroshi
no journal, ,
no abstracts in English
Takahatake, Yoko; Ambai, Hiromu; Sano, Yuichi; Takeuchi, Masayuki; Watanabe, Masayuki; Sakamoto, Kan*; Yamashita, Shinichiro
no journal, ,
no abstracts in English
Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Tachi, Yoshiaki; Otsuka, Satoshi; Kaito, Takeji
no journal, ,
9Cr-ODS steel was irradiated to approximately 240 dpa at 700 C by using an ion irradiation where the irradiation condition is close to that of a commercial fast reactor cladding tube. Microstructural observation was performed on the nano-sized particles, which is the main factors of the excellent high-temperature strength of 9Cr-ODS steel. The observation revealed that the nano-sized articles of about 5 nm were uniformly dispersed with high density even after irradiation up to 240 dpa. Furthermore, no void formation was observed. Thus, it was concluded that the nano-sized articles of 9Cr-ODS steel exist stably under a high-dose irradiation condition at a practical level at 700 C.
Tanno, Takashi; Yano, Yasuhide; Oka, Hiroshi*
no journal, ,
3-dimensional atom prove tomography (3D-APT) was applied to JAEA 9Cr-ODS steel as trial, and the dispersion condition of nano-sized oxide was evaluated. 3D-APT was successfully carried out and the dispersion condition in tiny tip was obtained. 3D-APT mapping showed that the nano-sized oxide is Y-Ti-O complexed, and there could be chemical distribution.
Tanno, Takashi; Oka, Hiroshi*; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji
no journal, ,
Fe + He dual ions beam irradiation to ODS steels and other materials which are developed as the candidates for fuel cladding tubes of fast reactor was carried out in order to evaluate their swelling resistance. Austenitic steels and tempered martensitic steels, which are not ODS, were swelled due to coarse voids formation by the irradiation. On the other hand, the swelling of ODS steels was much smaller than the others after irradiation up to 95 dpa. The fact shows that the fine dispersed oxide in ODS steels can suppress the swelling strongly.
Oka, Hiroshi*; Hashimoto, Naoyuki*; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji
no journal, ,
The effect of nitrogen concentration on microstructure and high temperature creep strength of 9Cr-ODS ferritic/martensitic steel was investigated. The results showed that the tensile strength and high temperature creep strength decreased with increasing nitrogen concentration. The microstructural analysis suggests that the non-uniform dispersion of Y-Ti-O particles due to the increase in nitrogen concentration may have been the cause of the decrease in high temperature creep strength. The non-uniform dispersion of Y-Ti-O particles is probably the result of the consumption of Ti by the formation of Ti nitrides, which weakened its refinement effect on the nano-sized oxide particles.
Toyama, Takeshi*; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Nakashima, Hideharu*; Onuma, Masato*
no journal, ,
no abstracts in English
Mitsuhara, Masatoshi*; Kurino, Koichi*; Nakashima, Hideharu*; Yano, Yasuhide; Otsuka, Satoshi; Onuma, Masato*; Toyama, Takeshi*
no journal, ,
no abstracts in English
Yamazaki, Jin*; Onuma, Masato*; Otsuka, Satoshi; Tanno, Takashi; Toyama, Takeshi*; Mitsuhara, Masatoshi*; Nakashima, Hideharu*
no journal, ,
X-ray small angle scattering (SAXS) analysis was carried out in order to evaluate the stability of nano-sized dispersed oxide particles in oxide dispersed strengthened (ODS) steels after creep test. The size of oxide particles was not changed after creep tests at 800C and below. On the other hand, the size tended to increase in the sampled tested at over 800C. However, the SAXS analysis showed that the nano-sized YTiO
oxide particle still existed in the sample after the creep test at 1000C. The result shows that dispersed nano-sized oxide particles in ODS steels have superior stability under creep test at high temperature.
