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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
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.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Abe, Yutaka*; Tsubakihara, Kosuke*; Okumura, Shin*; Ishizuka, Chikako*; Yoshida, Tadashi*; et al.
Journal of Nuclear Science and Technology, 60(1), p.1 - 60, 2023/01
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：1 Percentile：71.47(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.
Otsuka, Satoshi; Shizukawa, Yuta; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Onizawa, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.
Journal of Nuclear Science and Technology, 11 Pages, 2022/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)JAEA has been developing 9Cr-oxide dispersion strengthened (ODS) tempered martensitic steel(TMS) as a candidate material for the fuel cladding tubes of sodium-cooled fast reactors(SFRs). The reliable prediction of in-reactor creep-rupture strength is critical for implementing the 9Cr-ODS TMS cladding tube in the SFR. This study investigated the quantitative correlation between the creep properties of 9Cr-ODS TMS at 700 C and the dispersions of nanosized oxides by analyzing the creep data and the material's nanostructure. The possibility of deriving a formula for estimating the in-reactor creep properties of 9Cr-ODS TMSs based on an analysis of the nanostructure of neutron-irradiated 9Cr-ODS TMSs was also discussed. The creep properties of 9Cr-ODS TMS at 700 C closely correlated with the dispersion of nanosized oxide particles. The correlation between creep-rupture lives and nanosized oxide particle dispersion was determined using existing creep models. The elucidation of correlation between the stress exponent of secondary creep rate and the nanostructure is essential to enhance future modeling reliability and formulation.
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Materials, 555, p.153105_1 - 153105_8, 2021/11
Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)The aim of this study was to evaluate the tensile properties and microstructures of dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging at temperatures between 400 and 600C up to 30,000 h. Characterization of microstructure was carried out by scanning electron microscopy and transmission electron microscopy. Microstructural analysis showed that the microstructure in the weld metals consisted of lath martensite containing a small amount of residual austenite. Thermal aging hardening of WMs occurred at 400 and 450C due to the effects of both a-a' phase separation and G-phase precipitation. However, there was no significant change in the total elongation, and fracture surfaces indicated that very fine dimpled rupture was predominant rather than the cleavage rupture. It was suggested that lath martensite phases enhanced the tensile strength due to phase separation, while residual austenite played a role in keeping elongation as a soft phase.
Villaret, F.*; Boulnat, X.*; Aubry, P.*; Yano, Yasuhide; Otsuka, Satoshi; Fabregue, D.*; de Carlan, Y.*
Materials Science & Engineering A, 824, p.141794_1 - 141794_10, 2021/09
Times Cited Count：1 Percentile：17.89(Nanoscience & Nanotechnology)Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Kaito, Takeji; Ukai, Shigeharu*
Materials Transactions, 62(8), p.1239 - 1246, 2021/08
Times Cited Count：0 Percentile：0(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.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki
Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04
Times Cited Count：4 Percentile：78.24(Materials Science, Multidisciplinary)In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.
Ar by the (
,2) reactionJuh
sz, M. M.*; Elekes, Z.*; Sohler, D.*; Utsuno, Yutaka; Yoshida, Kazuki; Otsuka, Takaharu*; Ogata, Kazuyuki*; Doornenbal, P.*; Obertelli, A.*; Baba, Hidetada*; et al.
Physics Letters B, 814, p.136108_1 - 136108_8, 2021/03
Times Cited Count：2 Percentile：32.08(Astronomy & Astrophysics)The nuclear structure of Ar was studied by the (,2) reaction using -ray spectroscopy for the bound and unbound states. Comparing the results to our shell-model calculations, two bound and six unbound states were established. The low cross sections populating the two bound states of Ar could be interpreted as a clear signature for the presence of significant sub-shell closures at neutron numbers 32 and 34 in argon isotopes.
Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi
Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.
Ukai, Shigeharu; Kato, Shoichi; Furukawa, Tomohiro; Otsuka, Satoshi
Materials Science & Engineering A, 794, p.139863_1 - 139863_13, 2020/09
Times Cited Count：23 Percentile：91.97(Nanoscience & Nanotechnology)The FeCrAl-oxide dispersion strengthened (ODS) alloy is the promising cladding material for the accident-tolerant fuel (ATF) of the light water reactors (LWR). Ring-creep tests for FeCrAl-ODS alloy cladding were carried out at 973 K and 1273 K. The dislocation detachment stress from the dispersoid was derived by considering the dislocation-dispersoid elastic interaction and the dislocation relaxation effect by climb motion. When the applied stress exceeds the dislocation detachment stress, dislocations overcome the dispersoids with the reduced values of the stress exponent. When the stress is lower than the dislocation detachment stress, grain boundary sliding (GBS) is dominant factor for the low strain rate creep deformation at 1273 K. Based on those findings, new constitutive equations for creep deformation were constructed, which is applicable to low stress, low strain rate and high temperature conditions encountered at the reactor sever accident.
Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi
Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08
Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Ca corroborates arising 
neutron magic numberChen, S.*; Lee, J.*; Doornenbal, P.*; Obertelli, A.*; Barbieri, C.*; Chazono, Yoshiki*; Navrtil, P.*; Ogata, Kazuyuki*; Otsuka, Takaharu*; Raimondi, F.*; et al.
Physical Review Letters, 123(14), p.142501_1 - 142501_7, 2019/10
Times Cited Count：37 Percentile：92.75(Physics, Multidisciplinary)no abstracts in English
Yano, Yasuhide; Sekio, Yoshihiro; Tanno, Takashi; Kato, Shoichi; Inoue, Toshihiko; Oka, Hiroshi; Otsuka, Satoshi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; et al.
Journal of Nuclear Materials, 516, p.347 - 353, 2019/04
Times Cited Count：11 Percentile：82.95(Materials Science, Multidisciplinary)9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 C, which surpassed that of heat-resistant austenitic steels although creep rupture strength of tempered martensitic steels is generally lower than that of austenitic steels at high temperatures. The measured creep rupture strength of 9Cr-ODS steel claddings at 1000 C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 C. This superior strength seemed to be owing to transformation of the matrix from the -phase to the -phase. The transient burst strengths for 9Cr-ODS steel were much higher than those for 11Cr-ferritic/martensitic steel (PNC-FMS). Cumulative damage fraction analyses suggested that the life fraction rule can be used for the rupture life prediction of 9Cr-ODS steel and PNC-FMS claddings in the transient and accidental events with a certain accuracy.
Otsuka, Satoshi; Kaito, Takeji
Enerugi Rebyu, 39(1), p.44 - 46, 2019/01
For performance improvement of next-generation nuclear system such as fast reactor, it has been expected to develop advanced material resistant to severe in-reactor environment (i.e. high-dose neutron irradiation at high-temperature). Japan Atomic Energy Agency (JAEA) has been developing Oxide Dispersion Strengthened (ODS) ferritic steel for long life fuel cladding tube of fast reactor. Application of ODS ferritic steel to fast reactor fuel can extend the fuel life time twice or more as long as the fuel with conventional cladding tube (i.e. modified SUS316), thus reducing fuel exchange frequency and fuel cost. It can be adaptable to high-temperature plant operation, which is favorable for improvement of power generation efficiency. This paper interprets the development of ODS ferritic steel cladding tube for sodium-cooled fast reactor, which has been led by JAEA for dozens of years.
Ne by proton knockout reactionsMurray, I.*; MacCormick, M.*; Bazin, D.*; Doornenbal, P.*; Aoi, Nori*; Baba, Hidetada*; Crawford, H. L.*; Fallon, P.*; Li, K.*; Lee, J.*; et al.
Physical Review C, 99(1), p.011302_1 - 011302_7, 2019/01
Times Cited Count：14 Percentile：86.81(Physics, Nuclear)no abstracts in English
Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji
Nuclear Materials and Energy (Internet), 16, p.230 - 237, 2018/08
Times Cited Count：4 Percentile：44.5(Nuclear Science & Technology)