Researcher Database

Hiroshi Oka
Faculty of Engineering Materials Science and Engineering Energy Materials
Assistant Professor

Researcher Profile and Settings

Affiliation

  • Faculty of Engineering Materials Science and Engineering Energy Materials

Job Title

  • Assistant Professor

Degree

  • Doctor of Philosophy(2014/03 Hokkaido University)

J-Global ID

Research Interests

  • Oxide Dispersion Strengthened steel   Transmission Electron Microscopy   

Research Areas

  • Nanotechnology/Materials / Metallic materials / Nuclear Material

Academic & Professional Experience

  • 2020/04 - Today Hokkaido University Faculty of Engineering Division of Materials Science and Engineering Assistant Professor
  • 2018/04 - 2020/03 Japan Atomic Energy Agency Fuels and Materials Department Research Engineer
  • 2014/04 - 2018/03 Japan Atomic Energy Agency Fast Reactor Fuel Cycle Technology Development Department Research Engineer

Education

  • 2011/04 - 2014/03  Hokkaido University  Graduate School of Engineering  Division of Materials Science and Engineering

Association Memberships

  • ATOMIC ENERGY SOCIETY OF JAPAN   THE JAPAN INSTITUTE OF METALS AND MATERIALS   Materials Research Society   

Research Activities

Published Papers

  • Y. Yano, Y. Sekio, T. Tanno, S. Kato, T. Inoue, H. Oka, S. Ohtsuka, T. Furukawa, T. Uwaba, T. Kaito, S. Ukai
    Journal of Nuclear Materials 516 347 - 353 0022-3115 2019/04 [Refereed][Not invited]
  • Development of ODS tempered martensitic steel for high burn up fuel cladding tube of SFR
    S. Ohtsuka, T. Tanno, H. Oka, Y. Yano, Y. Tachi, T. Kaito, R. Hashidate, S. Kato, T. Furukawa, C. Ito, T. Yoshitake
    Proceedings of GIF symposium 2018 2019 [Refereed][Not invited]
  • Hiroshi Oka, Takashi Tanno, Satoshi Ohtsuka, Yasuhide Yano, Takeji Kaito
    Nuclear Materials and Energy 16 230 - 237 2018/08 [Refereed][Not invited]
     
    © 2018 The objective of this study was to investigate the effect of nitrogen concentration on mechanical properties and nano-structure of 9Cr oxide dispersion strengthened (ODS) ferritic/martensitic steel. 9Cr-ODS specimens with the wide range of nitrogen concentration, from 0.004 to 0.110 wt%, were systematically investigated by hardness and tensile tests and several microstructural characterization methods. Hardness and tensile strength at 973 K were significantly decreased as nitrogen concentration increased, due to the decrease in the amount of the residual α-ferrite phase. Coarse inclusions containing Y and Ti, which could negatively affect creep strength and processability, were formed, and that suggested degradation of the nano-particle distribution. The technical knowledge obtained in this study will contribute towards the setting of a reasonable nitrogen concentration specification for 9Cr-ODS steel.
  • Satoshi Ohtsuka, Takashi Tanno, Hiroshi Oka, Yasuhide Yano, Shoichi Kato, Tomohiro Furukawa, Takeji Kaito
    Journal of Nuclear Materials 505 44 - 53 0022-3115 2018/07/01 [Refereed][Not invited]
     
    A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in inlet sodium bulk flow was set at 0.07 wt ppm in the calculation.
  • Satoshi Ohtsuka, Takashi Tanno, Hiroshi Oka, Yasuhide Yano, Shoichi Kato, Tomohiro Furukawa, Takeji Kaito
    Journal of Nuclear Materials 0022-3115 2018/01/01 [Not refereed][Not invited]
  • Y. Yano, T. Tanno, H. Oka, S. Ohtsuka, T. Inoue, S. Kato, T. Furukawa, T. Uwaba, T. Kaito, S. Ukai, N. Oono, A. Kimura, S. Hayashi, T. Torimaru
    Journal of Nuclear Materials 487 229 - 237 0022-3115 2017/04 [Refereed][Not invited]
     
    © 2017 Elsevier B.V. Ultra-high temperature ring tensile tests were performed to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions with temperatures ranging from room temperature to 1400 °C which is close to the melting point of core materials. The experimental results showed that the tensile strength of 9Cr-ODS steel claddings was highest in the core materials at ultra-high temperatures of 900–1200 °C, but there was significant degradation in the tensile strength of 9Cr-ODS steel claddings above 1200 °C. This degradation was attributed to grain boundary sliding deformation with γ/δ transformation, which is associated with reduced ductility. By contrast, the tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 °C, unlike the other tested materials.
  • Tomoyuki Uwaba, Yasuhide Yano, Satoshi Ohtsuka, Masayuki Naganuma, Takashi Tanno, Hiroshi Oka, Shoichi Kato, Takeji Kaito, Shigeharu Ukai, Akihiko Kimura, Shigenari Hayashi, Tadahiko Torimaru
    2017 International Congress on Advances in Nuclear Power Plants, ICAPP 2017 - A New Paradigm in Nuclear Power Safety, Proceedings 2017/01/01 [Refereed][Not invited]
     
    Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system is actuated.
  • Takashi Tanno, Yasuhide Yano, Hiroshi Oka, Satoshi Ohtsuka, Tomoyuki Uwaba, Takeji Kaito
    Nuclear Materials and Energy 9 353 - 359 2352-1791 2016/12/01 [Not refereed][Not invited]
     
    Materials for core components of fusion reactors and fast reactors, such as blankets and fuel cladding tubes, must offer the best possible high temperature strength and irradiation resistance because they will be exposed to high heat flux and heavy neutron irradiation. Japan Atomic Energy Agency (JAEA) has been developing 9 and 11 chromium (Cr) oxide dispersion strengthened (ODS) steels as candidate materials for advanced fast reactor cladding tubes. In this study the JAEA 11Cr-ODS steel was rolled in order to evaluate its anisotropy. The tensile tests and creep tests were carried out at 700 °C in longitudinal and transverse orientations. The anisotropy of the tensile strength was negligible, though that of the creep strength was distinct. The observation results and chemical composition mapping suggested that the cause of the anisotropy in the creep strength was a previously formed columnar boundary, that is, a prior particle boundary including Ti-rich sub-micro metric precipitates.
  • Y. Yano, T. Tanno, Y. Sekio, H. Oka, S. Ohtsuka, T. Uwaba, T. Kaito
    Nuclear Materials and Energy 9 324 - 330 2352-1791 2016/12/01 [Not refereed][Not invited]
     
    The relationship among tensile strength, Vickers hardness and dislocation density for two types of 11Cr-ferritic/martensitic steel (PNC-FMS) was investigated after aging at temperatures between 400 and 800 °C up to 45,000 h and after neutron irradiation. A correlation between tensile strength and Vickers hardness was expressed empirically. The linear relationship for PNC-FMS wrapper material was observed between yield stress and the square of dislocation density at RT and aging temperature according to Bailey–Hirsch relation. Therefore, it was clarified that the correlation among dislocation density, tensile strength and Vickers hardness to aging temperature was in good agreement. On the other hand, the relationship between tensile strength ratio when materials were tested at aging temperature and Larson–Miller parameter was also in excellent agreement with aging data between 400 and 700 °C. It was suggested that this correlation could use quantitatively for separately evaluating irradiation effects from neutron irradiation data containing both irradiation and aging effects.
  • Keisuke Takahashi, Hiroshi Oka, Somei Ohnuki
    ACS APPLIED MATERIALS & INTERFACES 8 (6) 3725 - 3729 1944-8244 2016/02 [Not refereed][Not invited]
     
    The locations and roles of noble gases at an oxide/metal interface in oxide dispersed metal are theoretically and experimentally investigated. Oxide dispersed metal consisting of FCC Fe and Y2Hf2O7 (Y2Ti2O7) is synthesized by mechanical alloying under a saturated Ar gas environment. Transmission electron microscopy and density functional theory observes the strain field at the interface of FCC Fe {111} and Y2Hf2O7 {111} whose physical origin emerges from surface reconstruction due to charge transfer. Noble gases are experimentally observed at the oxide (Y2Ti2O7) site and calculations reveal that the noble gases segregate the interface and grow toward the oxide site. In general, the interface is defined as the trapping site for noble gases; however, transmission electron microscopy and density functional theory found evidence which shows that noble gases grow toward the oxide, contrary to the generally held idea that the interface is the final trapping site for noble gases. Furthermore, calculations show that the inclusion of He/Ar hardens the oxide, suggesting that material fractures could begin from the noble gas bubble within the oxides. Thus, experimental and theoretical results demonstrate that noble gases grow from the interface toward the oxide and that oxides behave as a trapping site for noble gases.
  • Hiroshi Oka, Yutaka Sato, Naoyuki Hashimoto, Somei Ohnuki
    JOURNAL OF NUCLEAR MATERIALS 462 470 - 474 0022-3115 2015/07 [Not refereed][Not invited]
     
    Depth dependence of hardness in ion-irradiated 316 stainless steel was evaluated by the sectioning of damaged region and subsequent nano-indentation. The range of plastically deformed region by the nano-indentation was supplementarily investigated by transmission electron microscopy. When the indentation depth was the critical indentation depth, h(c) which was derived from the Nix-Gao plot, the dislocation structure was observed from the specimen surface to right below the bottom of the ion-irradiated region. To verify the depth dependence of hardness, the "multi-layer model" was introduced in this study. The multi-layer model is based on the following assumptions; (1) the ion-irradiated region can be divided into sub-layers having their own local hardness, H-L; (2) the hardness can be the product off and H-L in each sub-layer where f is the volume fraction of a deformation zone; and (3) the deformation zone can be a hemisphere. Eventually, through the sectioning and following nano-indentation, H-L, in each sub-layer was experimentally evaluated. Further the correlation between the displacement damage and the irradiation hardening Delta H in this study agreed with that of neutron irradiation experiments. (C) 2015 Elsevier B.V. All rights reserved.
  • Multi-layer Method combined with Nano-indentation, FIB and XTEM for Nano-hardness Measurement
    R. Kurishiba, T. Endo, N. Miyazaki, Y. Wang, H. Oka, Y. Sato, A. Sawa, N. Hashimoto, S. Ohnuki
    MICROSCOPY Oxford 64 I119  2015 [Refereed][Not invited]
  • H. Oka, N. Hashimoto, T. Muroga, A. Kimura, M. A. Sokolov, T. Yamamoto, S. Ohnuki
    JOURNAL OF NUCLEAR MATERIALS 455 (1-3) 454 - 459 0022-3115 2014/12 [Not refereed][Not invited]
     
    F82H-IEA and its EB-weld joint were irradiated at 573 and 773 K up to 9.6 dpa and the irradiation effect on its mechanical properties and microstructure were investigated. A hardness profile across the weld joint before irradiation showed the hardness in transformed region (TR) was high and especially that in the-edge of TR was the highest (high hardness region: HHR) compared to base metal (BM). These hardness distribution was correspond to grain size distribution. After irradiation, hardening in HHR was small compared to other region in the sample. In tensile test, the amount of hardening in yield strength and ultimate tensile strength of F82H EB-weld joint was almost similar to that of F82H-IEA but the fracture position of EB-weld joint was at the boundary of TR and BM. Therefore, the TR/BM boundary is the structural weak point in F82H EB-weld joint after irradiation. As the plastic instability was observed, the dislocation channeling deformation can be expected though the dislocation channel was not observed in this study. (C) 2014 Elsevier B.V. All rights reserved.
  • Hiroshi Oka, Masashi Watanabe, Somei Ohnuki, Naoyuki Hashimoto, Shinichiro Yamashita, Satoshi Ohtsuka
    JOURNAL OF NUCLEAR MATERIALS 447 (1-3) 248 - 253 0022-3115 2014/04 [Not refereed][Not invited]
     
    An oxide dispersion strengthened (ODS) austenitic stainless steel was developed by mechanical alloying (MA) of advanced SUS316 stainless steel. A nano-characterization was performed to understand details of the effect of minor alloying elements in the distribution of dispersoids. It is shown that Y2O3 particles dissolve into the austenitic matrix after the MA for 6 h. Annealing at 1073 K or higher temperatures result in a distribution of fine oxide particles in the recrystallized grains in the ODS austenitic stainless steel. Additions of Hafnium or Zirconium led to the distribution of finer oxide particles than in samples without these elements, resulting in an increase in the hardness of the samples. The most effective concentration of Hf and Zr to increase the hardness was 0.6 and 0.2-0.3 wt%, respectively. (C) 2014 Elsevier B.V. All rights reserved.
  • Tengfei Zhang, Shigehito Isobe, Yongming Wang, Hiroshi Oka, Naoyuki Hashimoto, Somei Ohnuki
    JOURNAL OF MATERIALS CHEMISTRY A 2 (12) 4361 - 4365 2050-7488 2014 [Not refereed][Not invited]
     
    In this study, LiTi2O4 was synthesized as a possible catalyst for complex metal hydrides. LiTi2O4 was stable in the sample after high-energy ball milling and heat treatment. LiTi2O4 exhibited a catalytic effect among the samples of MgH2, LiAlH4 and LiNH2. The desorption kinetics and the purity of the desorbed hydrogen gas have been improved by doping LiTi2O4. Furthermore, the catalytic mechanism of LiTi2O4 was discussed in accordance with the experimental results.
  • Naoyuki Hashimoto, Hiroshi Oka, Takeo Muroga, Takuya Nagasaka, Akihiko Kimura, Shigeharu Ukai, Takuya Yamamoto, Michail A. Sokolov
    MATERIALS TRANSACTIONS 54 (4) 442 - 445 1345-9678 2013 [Not refereed][Not invited]
     
    Under the TITAN project, in order to determine the contributions of different microstructural features to strength and to deformation mode, microstructure of deformed flat tensile specimens of irradiated reduced activation F82H IEA and its joint were investigated by transmission electron microscopy (TEM), following tensile test and fracture surface examination by scanning electron microscopy (SEM). After irradiation, changes in yield strength, deformation mode, and strain-hardening capacity were seen, with the magnitude of the changes dependent on irradiation temperature. Irradiation to F82H IEA at 573 K led to a significant loss of strain-hardening capacity with a large change in yield strength. There was a tendency for a reduction in strain rate to cause a decrease in yield strength and elongation. While, irradiation at 773 K had little effect on strength, but a reduction in strain rate caused a decrease in ductility. SEM revealed fracture surfaces showing a martensitic mixed quasi-cleavage and ductile-dimple fracture in all samples. TEM have exhibited defect free bands (dislocation channels) in the necked region irradiated at 573 K. This suggests that dislocation channeling would be the dominant deformation mechanism in this steel irradiated at 573 K, resulting in the loss of strain-hardening capacity. While, the necked region of the irradiated F82H IEA joint, where showed less hardening than F82H IEA, has showed deformation bands only. From these results, it is suggested that the pre-irradiation microstructure, especially the dislocation density, could affect the post-irradiation deformation mode.
  • Hiroshi Oka, Masashi Watanabe, Naoyuki Hashimoto, Somei Ohnuki, Shinichiro Yamashita, Satoshi Ohtsuka
    Journal of Nuclear Materials 442 (1-3) 164 - 168 0022-3115 2013 [Not refereed][Not invited]
     
    In this study, identification of the crystal structure and analysis of the orientation relationship of oxide particles in an oxide dispersion strengthened austenitic stainless steel was carried out. High resolution transmission electron microscopy (HRTEM) and energy dispersive spectroscopy showed that most of the oxide particles had a faceted shape and consisted of a complex oxide, the anion-deficient fluorite structure Y2Hf2O7. Selected area diffraction patterns and HRTEM indicated that the faceted oxide particle has a cube-on-cube orientation relationship with the surrounding matrix. In addition, strain fields were observed around the oxide particle with given reflection conditions, indicating that it surrounds the oxide particle. The observed strain fields would affect glide dislocation pinning and the migration of irradiation-induced point defects. © 2013 Elsevier B.V. All rights reserved.
  • C. Z. Yu, H. Oka, N. Hashimoto, S. Ohnuki
    JOURNAL OF NUCLEAR MATERIALS 417 (1-3) 286 - 288 0022-3115 2011/10 [Not refereed][Not invited]
     
    Development of irradiation damage structure in ODS ferritic steels was studied by means of in-situ observation of HVEM up to 10 dpa at 773 K. In this study, standard ODS steel, Fe-16Cr-4A1-2W-0.35Y(2)O(3) and other of steels including Zr or Hf were examined. During electron-irradiation, preferential nucleation of dislocation loops was observed on oxide particles. In Hf- or Zr-added steels dislocation loops showed a low growth rate and a low number density compared with the standard steel. These results are explained by the sink effect at the interface between the matrix and oxide particles. (C) 2011 Elsevier B.V. All rights reserved.
  • H. Oka, M. Watanabe, H. Kinoshita, T. Shibayama, N. Hashimoto, S. Ohnuki, S. Yamashita, S. Ohtsuka
    JOURNAL OF NUCLEAR MATERIALS 417 (1-3) 279 - 282 0022-3115 2011/10 [Not refereed][Not invited]
     
    The oxide dispersion strengthening method was applied to an austenitic stainless steel based on SUS316 by mechanical alloying with additions of minor alloying elements. This ODS austenitic stainless steel was electron-irradiated to investigate the damage structure. Microstructural observation revealed that ODS austenitic stainless steel has a fine distribution of complex oxides. The in situ observation during electron irradiation showed that both dislocation loops and small cavities nucleated at the interface between oxide particles and matrix. In the case of helium implantation, defect clusters also nucleated at the interface. These results suggest that the oxide interface is an effective sink for irradiation-induced point defects and helium atoms. (C) 2010 Elsevier B.V. All rights reserved.
  • Hiroshi Oka, Yosuke Yamazaki, Hiroshi Kinoshita, Naoyuki Hashimoto, Somei Ohnuki, Shinichiro Yamashita, Satoshi Ohtsuka
    Materials Research Society Symposium Proceedings 1298 21 - 25 0272-9172 2011 [Not refereed][Not invited]
     
    Oxide dispersion strengthened austenitic stainless steel (ODS316), which is based on advanced SUS316 steel, has been developed by mechanically alloying and hot extrusion. Hafnium and titanium were added to make a fine distribution of oxide particles. The stability of oxide particles dispersed in ODS316 under irradiation was evaluated after 250 keV Fe+ irradiation up to high doses at 500°C. TEM observation and EDS analysis indicated that fine complex oxide particles with Y, Hf and Ti were mainly dispersed in the matrix. There are no significant changes in the distribution and the size of oxide particles after irradiation. It was also revealed that the constitution ratio of Ti in complex oxide appeared to be decreased after irradiation. This diffuse-out of Ti during irradiation could be explained by the difference in oxide formation energy among alloying elements. © 2011 Materials Research Society.

Conference Activities & Talks

  • Mass production technology development of 9Cr-ODS steel; Development of Prototype large ATTRItor for mass production of ODS steel (PATTRIODS) and test production
    Oka, Hiroshi, Tanno, Takashi, Yano, Yasuhide, Otsuka, Satoshi, Kaito, Takeji, Tachi, Yoshiaki
    日本金属学会2020年春期講演大会  2020/03
  • Stability of nano-particles in ODS steel cladding for fast reactor irradiated to about 240 dpa
    Oka, Hiroshi, Tanno, Takashi, Yano, Yasuhide, Tachi, Yoshiaki, Otsuka, Satoshi, Kaito, Takeji
    QST高崎サイエンスフェスタ2019  2019/12
  • Microstructural stability of ODS steel after long-time creep test
    Hiroshi Oka, Takashi Tanno, Yasuhide Yano, Satoshi Ohtsuka, Takeji Kaito, Yoshiaki Tachi
    the 19th International Conference on Fusion Reactor Materials (ICFRM-19)  2019/10
  • Tensile properties on dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging
    Yano, Yasuhide, Tanno, Takashi, Oka, Hiroshi, Sekio, Yoshihiro, Otsuka, Satoshi, Kaito, Takeji, Tachi, Yoshiaki
    the 19th International Conference on Fusion Reactor Materials (ICFRM-19)  2019/10
  • Tensile property changes of 11Cr ferritic/martensitic steel irradiated in fast reactor Joyo
    Tanno, Takashi, Yano, Yasuhide, Oka, Hiroshi, Sekio, Yoshihiro, Otsuka, Satoshi, Kaito, Takeji, Tachi, Yoshiaki
    the 19th International Conference on Fusion Reactor Materials (ICFRM-19)  2019/10
  • Relationship between microstructure and manufacturing condition of 9Cr-ODS ferritic/martensitic steels  [Not invited]
    Hiroshi Oka, Takashi Tanno, Yasuhide Yano, Satoshi Ohtsuka, Takeji Kaito, Yoshiaki Tachi
    Fifth International Workshop on Structural Material for Innovative Nuclear Systems (SMINS-5)  2019/06
  • Model calculation of Cr dissolution from steel surface exposed to high-temperature flowing sodium
    Otsuka, Satoshi, Tanno, Takashi, Oka, Hiroshi, Yano, Yasuhide, Hashidate, Ryuta, Kato, Shoichi, Furukawa, Tomohiro, Kaito, Takeji, Ito, Chikara
    Fifth International Workshop on Structural Material for Innovative Nuclear Systems (SMINS-5)  2019/06
  • Outline of material irradiation research results using Joyo
    Kaito, Takeji, Yano, Yasuhide, Shizukawa, Yuta, Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi
    Fifth International Workshop on Structural Material for Innovative Nuclear Systems (SMINS-5)  2019/06
  • Post irradiation examinations for materials irradiated in Joyo
    Shizukawa, Yuta, Sekio, Yoshihiro, Oka, Hiroshi, Tanno, Takashi, Yano, Yasuhide, Tachi, Yoshiaki, Otsuka, Satoshi, Kaito, Takeji
    Fifth International Workshop on Structural Material for Innovative Nuclear Systems (SMINS-5)  2019/06
  • Tensile property changes of 11Cr ferritic/martensitic steel irradiated in Joyo
    Tanno, Takashi, Yano, Yasuhide, Sekio, Yoshihiro, Oka, Hiroshi, Otsuka, Satoshi, Kaito, Takeji, Tachi, Yoshiaki
    Fifth International Workshop on Structural Material for Innovative Nuclear Systems (SMINS-5)  2019/06
  • Effects of thermal aging on tensile properties of electron beam welded dissimilar joints between 11Cr-ferritic/martensitic steel and 316 stainless steel
    Yano, Yasuhide, Tanno, Takashi, Oka, Hiroshi, Sekio, Yoshihiro, Otsuka, Satoshi, Kaito, Takeji
    Fifth International Workshop on Structural Material for Innovative Nuclear Systems (SMINS-5)  2019/06
  • Evaluation of breach characteristics of fast reactor fuel pins during steady state irradiation  [Not invited]
    Hiroshi Oka, Yoshihisa Ikusawa, Satoshi Ohtsuka, Takeji Kaito
    NuMat 2018  2018/11
  • Cesium migration effects on irradiation behavior of fast reactor MOX fuel pins
    Tanno, Takashi, Oka, Hiroshi, Ikusawa, Yoshihisa, Uwaba, Tomoyuki, Otsuka, Satoshi, Kaito, Takeji, Maeda, Seiichiro
    Nuclear Materials Conference 2018 (NuMat 2018)  2018/10
  • Ultra-high temperature creep and transient burst strength of ODS steel cladding tube
    Yano, Yasuhide, Sekio, Yoshihiro, Kato, Shoichi, Tanno, Takashi, Inoue, Toshihiko, Oka, Hiroshi, Otsuka, Satoshi, Furukawa, Tomohiro, Uwaba, Tomoyuki, Kaito, Takeji, Ukai, Shigeharu
    Nuclear Materials Conference 2018 (NuMat 2018)  2018/10
  • Irradiation behavior of fast reactor fuel pins, 2; Irradiation behavior of annular pellet pins  [Not invited]
    Yokoyama, Keisuke, Uwaba, Tomoyuki, Tanno, Takashi, Oka, Hiroshi
    日本原子力学会2018年春の年会  2018/03  吹田 
    中空燃料は、燃料溶融に対する裕度の増加や燃料-被覆管機械的相互作用(FCMI)の緩和などの点で高性能化(高線出力化と高燃焼度化)に対して有効であると考えられており、高速実証炉と実用炉の高性能燃料として有望視されている。ここでは、中空燃料ピンの照射挙動を把握することを目的として、EBR-IIで定常照射されたMOX燃料ピンの照射後試験データ(燃料ピン外径測定、$\gamma$-スキャニング、金相試験)について中空燃料ピンと中実燃料ピンの比較評価を行った。MOX燃料ピンは中実燃料ピン及び中空燃料ピンの両方とも、ピーク燃焼度が40$\sim$130GWd/t、ピーク線出力が34$\sim$49kW/mの照射条件で照射された。評価の結果、中空燃料ピンは中実燃料ピンと比較して、FCMIが減少することにより燃料カラム部での外径増加が低減していることが、また、Cs移動が活発となり燃料カラム-ブランケット境界部にCsが蓄積しやすいために、境界部での局所的な外径増加が生じやすいことが示された。
  • Effect of nitrogen concentration on nano-structure and high-temperature strength of 9Cr-ODS steel  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi, Yano, Yasuhide, Kaito, Takeji
    18th International Conference on Fusion Reactor Materials (ICFRM-18)  2017/11  Aomori 
    In determining the nitrogen concentration specifications, nano-structure and high-temperature strength of 9Cr-ODS steel have been investigated as a function of the nitrogen content with the aim of obtaining technical knowledge that makes the specification reasonable. The hardness and tensile strength showed degradation with increasing nitrogen content. For a microstructure, the decrement of residual ferrite phase was confirmed. Since the nitrogen is an austenite stabilizer, the increment of nitrogen enhanced an alpha to $\gamma$ transformation, resulted in the decrease of the residual ferrite phase. It is considered that the reduction of the strength is due to the decrease of the residual ferrite phase.
  • Axial migration and accumulation behavior of cesium in fast reactor fuel pins, 2; Evaluation of diameter change of fuel pellet by cesium local accumulation  [Not invited]
    Tanno, Takashi, Oka, Hiroshi, Ikusawa, Yoshihisa, Uwaba, Tomoyuki, Kaito, Takeji
    日本原子力学会2017年秋の大会  2017/09  札幌 
    Evaluation of diameter changes of fuel pellets by cesium (Cs) accumulation based on gamma-scan profiles was carried out in order to estimate the influence of axial migration and accumulation behavior of Cs for fuel-cladding mechanical interaction (FCMI). Local diameter change of the UO$_{2}$ pellets facing MOX pellets and FCMI by the diameter change were predicted for EBR-II irradiation test. To prevent the FCMI, the radial gap should have been more 80$\mu$m larger than the initial radial gap as fabricated for EBR-II irradiation test.
  • Axial migration and accumulation behavior of cesium in fast reactor fuel pins, 3; Evaluation of cesium accumulation effect on lifetime of fuel cladding tube  [Not invited]
    Ikusawa, Yoshihisa, Uwaba, Tomoyuki, Tanno, Takashi, Oka, Hiroshi, Kaito, Takeji, Nemoto, Junichi
    日本原子力学会2017年秋の大会  2017/09  札幌 
    In high burnup MOX fuel pins, cesium accumulates at the UO$_{2}$ - MOX pellet boundary and cladding tube diameter is locally increased around the position by FCMI, due to the formation of Cs-U-O compound. Based on the computation results obtained using an irradiation behavior analysis code "CEDAR", we evaluated the effect of this behavior on cladding creep damage. As the computation result, it was found that the cladding creep damage around the UO$_{2}$ - MOX pellet boundary increases by the FCMI stress.
  • Axial migration and accumulation behavior of cesium in fast reactor fuel pins, 1; Axial migration and accumulation behavior of cesium observed in irradiation test of Joyo and foreign reactors  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Ikusawa, Yoshihisa, Uwaba, Tomoyuki, Kaito, Takeji
    日本原子力学会2017年秋の大会  2017/09  札幌 
    Through the irradiation test of Joyo and foreign reactors, the behavior of cesium migration and accumulation in axial heterogeneous fuel pins was investigated. The local increase of cladding tube diameter at MOX/UO$_{2}$ boundary was observed at a burnup of 115 GWd/t, which seems to be caused by the accumulation of cesium.
  • Influence of impurity nitrogen on microstructure and high-temperature strength of 9Cr-ODS steel  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi, Kaito, Takeji
    日本金属学会2017年秋期(第161回)講演大会  2017/09  札幌 
    9Cr-ODS鋼の優れた高温クリープ強度特性は同鋼が持つナノ組織に起因するものであるが、過去の研究では過剰酸素とTiの濃度変動がナノ組織とクリープ強度に大きな影響を及ぼすことがわかっている。一方で、同鋼の製造過程で混入する可能性のある不純物窒素の影響は調べられていない。そこで本研究では、不純物窒素が9Cr-ODS鋼の組織と強度に与える影響を明らかにするため、窒素含有量の異なる試料について微細組織解析および強度試験を行った。その結果、窒素濃度増加とともに硬さは低下し、残留$\alpha$フェライト相の割合は減少することが明らかとなった。強化因子である残留$\alpha$フェライト相の減少は、オーステナイト安定化元素である窒素の増加により$\alpha$$\rightarrow$$\gamma$変態の駆動力が増加したためであり、同相の減少は強度低下の主要因と考えられる。
  • Modelling and numerical calculation of mass transfer phenomena between fast reactor fuel cladding tube and liquid Na  [Not invited]
    Otsuka, Satoshi, Tanno, Takashi, Oka, Hiroshi, Yano, Yasuhide, Uwaba, Tomoyuki, Kaito, Takeji, Furukawa, Tomohiro, Kato, Shoichi
    日本金属学会2017年秋期(第161回)講演大会  2017/09  札幌 
    Maximum temperature of ODS steel cladding tube for long life fast reactor fuel is very high (approximately 700$^{\circ}$C) in normal operation condition. It was reported that, in reactor operation, mass transfer phenomena (dissolution, deposition, penetration) took place as a result of increased solubility of steel constituent elements in liquid Na. The driving force of these phenomena is the chemical potential gap of solute elements in steel and liquid Na, which is dependent of not only temperature but also other factors such as impurity concentrations in Liquid Na. For appropriately evaluating experimental data and predicting the corrosion behavior in actual plant, it is required to list up the key factors including other factors than temperature and residence time and understand the effects of these factors. In this study, transfer behavior of Cr (main alloying element of ODS steel) is discussed; modelling and numerical calculation were carried out on Cr dissolution behavior from fast reactor fuel cladding tube into liquid Na.
  • Development of ODS steel for fast reactor cladding material  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi, Yano, Yasuhide, Uwaba, Tomoyuki, Kaito, Takeji
    平成29年度日本原子力学会北関東支部若手研究者発表会  2017/04  東海 
    原子力機構では、高速炉用燃料被覆管の候補材として酸化物分散強化型(ODS: Oxide Dispersion Strengthened)鋼被覆管の開発を進めている。ODS鋼はナノサイズの酸化物粒子が母相に微細に分散した鉄鋼材料であり、優れた高温強度と耐照射性を有する。ODS鋼被覆管として、焼き戻しマルテンサイト相を主母相とするマルテンサイト系ODS鋼と再結晶フェライト組織を主母相とするフェライト系ODS鋼があるが、原子力機構では優れた耐照射性と製造性が期待できるマルテンサイト系ODS鋼を第一候補材として研究開発を進めている。マルテンサイト系ODS鋼の研究開発は、従来より9Cr-ODS鋼を中心に進めてきたが、近年、より耐食性を高めた11Cr-ODS鋼に関する研究開発も進められており、両鋼は類似の微細組織および高温強度特性を有することがわかっている。本発表ではマルテンサイト系ODS鋼に関して、均質性が高く高温強度の非常に優れた被覆管が製造可能な完全プレアロイ法の紹介や、将来的な量産体制確立に資する研究開発項目の一つとして、ODS鋼の組織発達に及ぼす加工熱処理の影響について述べる。
  • Evaluation on tolerance to failure of ODS ferritic steel claddings at the accident conditions of fast reactors  [Not invited]
    Uwaba, Tomoyuki, Yano, Yasuhide, Otsuka, Satoshi, Naganuma, Masayuki, Tanno, Takashi, Oka, Hiroshi, Kato, Shoichi, Kaito, Takeji, Ukai, Shigeharu, Kimura, Akihiko, Hayashi, Shigenari, Torimaru, Tadahiko
    2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017)  2017/04  Fukui/Kyoto 
    Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system operates.
  • R\&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 3-5; Evaluation on tolerance to failure of existing ODS ferritic steel claddings at the accident conditions of fast reactors  [Not invited]
    Uwaba, Tomoyuki, Yano, Yasuhide, Otsuka, Satoshi, Naganuma, Masayuki, Tanno, Takashi, Oka, Hiroshi, Kato, Shoichi, Kaito, Takeji, Ukai, Shigeharu, Kimura, Akihiko, Torimaru, Tadahiko, Hayashi, Shigenari
    日本原子力学会2017年春の年会  2017/03  平塚 
    事故時高温条件における燃料被覆管の破損限界評価は、高速炉の安全性を確保する上で極めて重要である。そのため、高速炉用9/12Cr-ODS鋼被覆管の1000$^{\circ}$Cまでの高温クリープ特性データを基に策定した破損寿命評価式を適用し、高速炉の事故時の被覆管の耐破損性を評価した。
  • Structure stability of ferritic ODS steel for fast reactor fuel cladding tube under irradiation  [Not invited]
    Tanno, Takashi, Oka, Hiroshi, Yano, Yasuhide, Otsuka, Satoshi, Kaito, Takeji
    放射線利用フォーラム2017 in 高崎 / 第1回QST高崎研シンポジウム  2017/01  高崎 
    Fe self-ion irradiation to ODS steels was conducted at 400$^{\circ}$C to evaluate the stability of oxide nano-dispersoids in the ODS steels and embrittlement behavior of higher Cr ODS steel under irradiation. Fe and He dual ion irradiation test at 470$^{\circ}$C was also conducted to evaluate the influence of He existence. The indentation hardness increased in early stage of the irradiation, and decreased over 150 dpa. But the hardness was higher than that as unirradiated, even if the dose reached 230 dpa. The Cr enrichment from 9Cr to 11Cr would not lead to extra irradiation hardening and/or irradiation embrittlement because the irradiation hardening behavior of 9Cr and 11Cr-ODS steels were almost same. The irradiation hardening due to Fe+He dual ions irradiation was negligible or comparatively small. Therefore it was considered that fine and dense voids formation enhanced by He existence was not significant.
  • Transient burst properties of ODS steel cladding for evaluating sever accident  [Not invited]
    Inoue, Toshihiko, Sekio, Yoshihiro, Otsuka, Satoshi, Yano, Yasuhide, Tanno, Takashi, Oka, Hiroshi, Furukawa, Tomohiro, Kaito, Takeji, Torimaru, Tadahiko, Hayashi, Shigenari, Kimura, Akihiko, Ukai, Shigeharu
    Nuclear Materials Conference 2016 (NuMat 2016)  2016/11  Montpellier 
    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.
  • High temperature creep properties of ODS steel cladding for evaluating severe accident  [Not invited]
    Kato, Shoichi, Furukawa, Tomohiro, Yano, Yasuhide, Tanno, Takashi, Otsuka, Satoshi, Oka, Hiroshi, Inoue, Toshihiko, Kaito, Takeji, Ukai, Shigeharu, Kimura, Akihiko, Hayashi, Shigenari, Torimaru, Tadahiko
    Nuclear Materials Conference 2016 (NuMat 2016)  2016/11  Montpellier 
    Oxide dispersion strengthened (ODS) steel is a prime candidate for cladding tubes of Japan Sodium-cooled Fast Reactor (JSFR) due to the high temperature and radiation resistances. One of the safety design of JSFR for Design Extension Condition (DEC) is the control of severe plant conditions, including prevention of severe accidents and mitigation of severe-accident consequences. Therefore, it is necessary to acquire the mechanical properties at ultra-high temperature conditions for core materials to evaluate safety design. There are, however, no data for ODS claddings at ultra-high temperature condition for the reflecting to the design criteria. In this study, creep rupture tests of 9Cr-ODS, 12Cr-ODS and FeCrAl-ODS steel claddings have been done at elevated temperatures, and the effect of minor elements such as Al, Zr and O on the mechanical strength and the creep rupture curve for the safety design were evaluated. The effect of minor elements was estimated based on the data at 700$^{\circ}$C and 1000$^{\circ}$C. As the results, it was confirmed that the addition of Zr had an effect on the improvement of creep strength at elevated temperature for the FeCrAl-ODS steel claddings.
  • Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions  [Not invited]
    Yano, Yasuhide, Tanno, Takashi, Oka, Hiroshi, Otsuka, Satoshi, Inoue, Toshihiko, Kato, Shoichi, Furukawa, Tomohiro, Uwaba, Tomoyuki, Kaito, Takeji, Ukai, Shigeharu, Ono, Naoko, Kimura, Akihiko, Hayashi, Shigenari, Torimaru, Tadahiko
    Nuclear Materials Conference 2016 (NuMat 2016)  2016/11  Montpellier 
    Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400$^{\circ}$C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200$^{\circ}$C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200$^{\circ}$C. This degradation was attributed to grain boundary sliding deformation with $\gamma$/$\delta$ transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 $^{\circ}$C unlike the other tested materials. Present study includes the result of "R\&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
  • Study on the physical properties of non-stoichiometric oxide fuels with high minor actinide contents, 4; Outline and main results  [Not invited]
    Tanaka, Kosuke, Seki, Takayuki, Oka, Hiroshi, Matsuda, Tetsushi, Muta, Hiroaki, Sekioka, Ken, Tokoro, Daishiro
    日本原子力学会2016年秋の大会  2016/09  久留米 
    The outline and main results of "Study on the physical properties of non-stoichiometric oxide fuels with high minor actinide contents", entrusted by the Ministry of Education, Culture, Sports, Science and Technology of Japan are presented.
  • R\&D of fuel cladding of ODS ferritic steel for maintaining fuel integrity at accidental high temperature condition, 2-1; Evaluation of failure limit correlation under an accident condition  [Not invited]
    Yano, Yasuhide, Kato, Shoichi, Otsuka, Satoshi, Inoue, Toshihiko, Tanno, Takashi, Oka, Hiroshi, Furukawa, Tomohiro, Kaito, Takeji, Kimura, Akihiko, Torimaru, Tadahiko, Hayashi, Shigenari, Ukai, Shigeharu
    日本原子力学会2016年春の年会  2016/03  仙台 
    事故時高温条件におけるODSフェライト鋼燃料被覆管の破損限界評価は、高速炉及び軽水炉の安全性を確保する上で極めて重要である。そのため、本公募で作製したAl含有高Cr-ODS鋼被覆管及び既存の高速炉用ODS鋼について、超高温における引張及びクリープ特性データ等を取得した。ここでは、超高温の各種強度試験結果について報告する。
  • Development of ODS steel cladding tube for fast reactor  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi, Yano, Yasuhide, Uwaba, Tomoyuki, Kaito, Takeji
    北海道大学大学院工学研究院材料科学部門記念シンポジウム「エネルギー社会における材料の役割; 最新の成果と展望」  2016/01  札幌 
    本発表では、原子力機構で実用化段階の高速炉用被覆管材料として開発が進められている酸化物分散強化型(ODS: Oxide Dispersion Strengthened)鋼について紹介する。ODS鋼はナノサイズの酸化物粒子が微細に分散した鉄鋼材料で、高温強度と耐照射性に優れる。焼き戻しマルテンサイト組織である9Cr-ODS鋼を中心に研究開発が進められ、単体元素粉末を製造プロセスから完全に排除した完全プレアロイ法の適用により、均質性が高く高温強度の非常に優れた被覆管が製造可能になった。また、腐食特性を向上させた11Cr-ODS鋼被覆管についても、9Cr-ODS鋼被覆管開発で得られた知見をベースに開発が進められている。加えて、将来的な量産体制確立に資する研究開発として、ODS鋼の組織発達に及ぼす加工熱処理の影響も検討されている。
  • Evaluation of irradiation resistance of ODS ferritic steel for fast reactor application  [Not invited]
    Tanno, Takashi, Oka, Hiroshi, Otsuka, Satoshi, Yano, Yasuhide, Kaito, Takeji
    第10回高崎量子応用研究シンポジウム  2015/10  高崎 
    原子力機構では高速炉の先進燃料被覆管候補材料として9Cr/11Cr-ODS鋼の開発を進めている。9Cr-ODS鋼については700$^{\circ}$C照射における酸化物粒子の照射下安定性を、Cr量の高い11Cr-ODS鋼については400$^{\circ}$C照射による$\alpha$'相形成とそれによる延性低下の有無を確認する必要がある。そこで本研究ではFeイオン照射を用いて短期間で上記特性の照射量依存性の傾向を得ることを目的とした。試料は完全プレアロイ法で作製した9Cr-ODS鋼と11Cr-ODS鋼の焼きならし・焼戻し(NT)材および徐冷(FC)材であり、400および700$^{\circ}$Cで最大150dpaまで照射し、微小押し込み硬さ試験にて(延性の指標として)照射硬化を評価した。400$^{\circ}$C照射では照射硬化は1GPa前後で飽和する傾向を示したものの、9Cr/11Cr-ODS鋼の挙動はほぼ同じであり、照射後延性に対するCr量増加の影響は有意でないと考えられる。700$^{\circ}$C照射では、9Cr/11Cr-ODS鋼ともに顕著な照射軟化はみられず、強度を担保する酸化物の分解や粗大化は80dpaまで起こっていないと考えられる。
  • Strength anisotropy of rolled 11Cr-ODS steel  [Not invited]
    Tanno, Takashi, Yano, Yasuhide, Oka, Hiroshi, Otsuka, Satoshi, Uwaba, Tomoyuki, Kaito, Takeji
    17th International Conference on Fusion Reactor Materials (ICFRM-17)  2015/10  Aachen 
    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 $^{\circ}$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.
  • Tensile properties and hardness of two types of 11Cr-ferritic/martensitic steel after aging up to 45,000 h  [Not invited]
    Yano, Yasuhide, Tanno, Takashi, Sekio, Yoshihiro, Oka, Hiroshi, Otsuka, Satoshi, Uwaba, Tomoyuki, Kaito, Takeji
    17th International Conference on Fusion Reactor Materials (ICFRM-17)  2015/10  Aachen 
    The relationship among tensile strength, Vickers hardness and dislocation density for two types of 11Cr-ferritic/martensitic steel (PNC-FMS) was investigated after aging at temperatures between 400 and 800$^{\circ}$C up to 45,000 h and after neutron irradiation. A correlation between tensile strength and Vickers hardness was expressed empirically. The linear relationship for PNC-FMS wrapper material was observed between yield stress and the square of dislocation density at RT and aging temperature according to Bailey-Hirsch relation. Therefore, it was clarified that the correlation among dislocation density, tensile strength and Vickers hardness to aging temperature to aging temperature was in good agreement. On the other hand, the relationship between tensile strength ratio when materials were tested at aging temperature and Larson-Miller parameter was also in excellent agreement with aging data between 400 and 700$^{\circ}$C. It was suggested that this correlation could use quantitatively for separately evaluating irradiation effects from neutron irradiation data containing both irradiation and aging effects.
  • Effect of thermo-mechanical treatments on nano-structure of 9Cr-ODS steel  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi, Yano, Yasuhide, Uwaba, Tomoyuki, Kaito, Takeji, Onuma, Masato
    17th International Conference on Fusion Reactor Materials (ICFRM-17)  2015/10  Aachen 
    The effect of thermo-mechanical treatments (TMTs) on the evolution of nano-structure in an oxide dispersion strengthened (ODS) ferritic/martensitic steel (Fe-9Cr-2W-0.22Ti-0.36Y2O3) was investigated. TMTs involve hot extruding and subsequent forging, which are expected to be part of a future industrial-scale manufacturing process of the ODS steel. It was shown that the ODS steel was composed of two phases - a fine-grained residual ferrite phase and a transformable martensite phase. The number density of the nano-sized particles in the residual ferrite phase was significantly higher than that in the martensite phase. The TMTs did not significantly affect the number density, but slightly affected the size distribution of the nano-sized particles in both ferrite phase and martensite phase. Moreover, the volume fraction of the residual ferrite phase decreased after TMTs. In summary, the TMT conditions could be a parameter which affects the nano-structure of the ODS steel.
  • R\&D of ODS ferritic steel cladding for maintaining fuel integrity at accident condition, 2; Hot-extruded bar and tube manufacturing test  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Inoue, Toshihiko, Otsuka, Satoshi, Yano, Yasuhide, Kaito, Takeji, Ukai, Shigeharu, Kimura, Akihiko, Torimaru, Tadahiko, Hayashi, Shigenari
    日本原子力学会2015年秋の大会  2015/09  静岡 
    事故耐性に優れる超高温用ODS鋼の開発に向けて、Cr濃度, Al濃度, Zr濃度, 酸素量をパラメータに、メカニカルアロイングと熱間押出により20鋼種の押出棒を製造した。10鋼種については薄肉被覆管とするため4回の冷間圧延試験を実施した。本研究は、文部科学省の原子力システム研究開発事業による委託業務として、北海道大学が実施した平成25-26年度「事故時高温条件での燃料健全性確保のためのODSフェライト鋼燃料被覆管の研究開発」の成果である。
  • Superior high-temperature strength of oxide dispersion strengthened (ODS) ferritic steel  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi, Yano, Yasuhide, Kaito, Takeji
    六ヶ所・核燃料サイクルセミナー; 核燃料サイクルの先端研究開発  2015/07  六ヶ所 
    経済性に優れる高速炉を実現するため、高温(700$^{\circ}$C), 高照射量(5$\times$10$^{27}$n/m$^{2}$, E$>$0.1MeV)の使用条件を満足する材料として酸化物分散強化型(ODS)フェライト鋼の開発を進めている。ODSフェライト鋼は耐中性子照射特性に優れるフェライト鋼中に、熱的に安定な酸化物粒子を分散させて高温強度の向上を図った材料であり、粉末冶金法により製造する。原子力機構で開発したODSフェライト鋼は、フェライト鋼としては世界最高レベルのクリープ強度を有しており、引張特性や疲労特性などにも優れる。
  • Effect of thermo-mechanical treatment on nano-structure of 9Cr-ODS steel  [Not invited]
    Oka, Hiroshi, Tanno, Takashi, Otsuka, Satoshi, Yano, Yasuhide, Kaito, Takeji, Onuma, Masato
    日本金属学会2015年春期(第156回)講演大会  2015/03  東京 
    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.
  • ヘリウム照射した316 鋼における照射硬化および照射後アニール硬化  [Not invited]
    唯木 康平, S. Chen, 岡 弘, 橋本 直幸, 大貫 惣明
    日本鉄鋼協会・日本金属学会両北海道支部合同冬季講演大会  2014/01  札幌
  • Hardness distribution and tensile properties in an electron beam weldment of F82H irradiated in HFIR  [Not invited]
    H. Oka, N. Hashimoto, T. Muroga, A. Kimura, M.A. Sokolov, T. Yamamoto
    16th International Conference on Fusion Reactor Materials (ICFRM-16)  2013/10  Beijin
  • Strength factor of dislocations in neutron irradiated austenitic stainless steel at high temperature  [Not invited]
    H. Oka, N. Hashimoto, S. Ohnuki, S. Yamashita
    16th International Conference on Fusion Reactor Materials (ICFRM-16)  2013/10  Beijin
  • オーステナイト系材料におけるHe照射硬化のナノインデンテーション法による評価  [Not invited]
    唯木 康平, 岡 弘, 陳 思維, 大貫 惣明
    日本金属学会2013年秋期(第153回)講演大会  2013/09  金沢大学
  • HFIR 照射した鉄系材料の損傷組織と引張特性に及ぼす水素の効果 (第一報)  [Not invited]
    岡 弘, 木村 友則, S. Wang, 橋本 直幸, 大貫 惣明
    日本金属学会2013年秋期(第153回)講演大会  2013/09  金沢大学
  • イオン照射したステンレス鋼のナノインデンテーション (セクショニング法の有用性)  [Not invited]
    岡 弘, 佐藤 豊, 橋本 直幸, 大貫 惣明
    日本金属学会2013年秋期(第153回)講演大会  2013/09  金沢大学
  • Evaluation of multi-layered hardness in ion-irradiated stainless steel by nano-indentation technique  [Not invited]
    Hiroshi Oka, Yutaka Sato, Naoyuki Hashimoto, Somei Ohnuki
    The International Symposium on Fusion Nuclear Technology (ISFNT-11)  2013/09  Barcelona
  • 長時間熱時効したオーステナイトステンレス鋼の微細組織と機械的特性  [Not invited]
    岡 弘, 橋本 直幸, 大貫 惣明, 山下真一郎
    平成25年度日本金属学会・日本鉄鋼協会両支部合同サマーセッション  2013/07  室蘭工業大学
  • FIBセクショニングによるナノ硬度分布の測定  [Not invited]
    岡 弘, 佐藤 豊
    第1回 北海道大学微小部・表面分析研究ユーザーズミーティング  2013/03  札幌
  • イオン照射したステンレス鋼のナノ硬さ評価法  [Not invited]
    佐藤 豊, 岡 弘, 大貫 惣明, 橋本 直幸
    日本鉄鋼協会・日本金属学会両北海道支部合同冬季講演大会  2013/01  室蘭工業大学
  • Hardening and Softening Caused by Long Term Neutron Irradiation in Modified -SUS316 Stainless Steel  [Not invited]
    H. Oka, T. Kubota, N. Hashimoto, S. Ohnuki, S. Yamashita
    2012 MRS fall meeting  2012/11  Boston
  • Evaluation by the nano indentation method of helium irradiation hardening in steel material  [Not invited]
    唯木 康平, 岡 弘, 陳 思維, 大貫 惣明, 橋本 直幸
    日本原子力学会2012年秋の大会  2012/09  広島
  • 中性子照射した改良SUS316鋼における組織と機械強度のミクロ−マクロ相関 (4)  [Not invited]
    岡 弘, 窪田 知宜, 橋本 直幸, 大貫 惣明, 山下 真一郎
    日本金属学会2012年秋期 (第151回)講演大会  2012/09  愛媛大学
  • Evaluation of helium induced hardening in steels by the nano indentation  [Not invited]
    K. Tadaki, H. Oka, S. Chen, S. Ohnuki, N. Hashimoto
    Asian-Core University Program on Advanced Energy Science; International Symposium on Advanced Energy Systems and Materials  2012/08  Aomori
  • Relationship between strength and microstructure of 316 steel at elevated temperature  [Not invited]
    H. Oka, N. Hashimoto, S. Ohnuki, S. Yamashita
    Asian-Core University Program on Advanced Energy Science; International Symposium on Advanced Energy Systems and Materials  2012/08  Aomori
  • オーステナイトODS鋼の酸化物粒子界面におけるHeの捕獲効果  [Not invited]
    唯木 康平, 岡 弘, 大貫 惣明, 橋本 直幸
    平成24年度日本金属学会・日本鉄鋼協会両支部合同サマーセッション  2012/07  北海道大学
  • 重照射した改良SUS316鋼における微細組織と機械強度の相関  [Not invited]
    岡 弘, 窪田 知宜, 橋本 直幸, 大貫 惣明, 山下 真一郎
    九州大学・応用力学研究所・炉内構造物の経年変化に関する研究集会  2012/07  九州大学
  • F82Hの接合・被覆部におけるはじき出し損傷と強度特性(日米協力TITAN計画)  [Not invited]
    岡 弘, 橋本 直幸, 室賀 健夫, 長坂琢也, 木村晃彦, 鵜飼重治, M.A. Sokolov, T. Yamamoto
    第9回核融合エネルギー連合講演会 -地球を救うエネルギー 核融合の未来-  2012/06  神戸
  • 中性子照射した改良SUS316鋼における組織と機械強度のミクロ−マクロ相関 (3)  [Not invited]
    山下 真一郎, 窪田 知宜, 岡 弘, 橋本 直幸, 大貫 惣明
    日本金属学会2012年春期 (第150回)講演大会  2012/03  横浜国立大学
  • 中性子照射した改良SUS316鋼における組織と機械強度のミクロ−マクロ相関 (2)  [Not invited]
    岡 弘, 窪田 知宜, 橋本 直幸, 大貫 惣明, 山下 真一郎
    日本金属学会2012年春期 (第150回)講演大会  2012/03  横浜国立大学
  • Interface structure of oxide particle in an ODS austenitic stainless steel  [Not invited]
    H.Oka, M.Watanabe, N.Hashimoto, S.Ohnuki, S.Yamashita, S.Ohtsuka
    Asian-Core University Program on Advanced Energy Science; International Symposium on Advanced Energy Systems and Materials  2012/01  Beijin
  • Microstructure-hardness correlation in ion-irradiated iron  [Not invited]
    B. Zhou, C.X. Liu, H. Oka, N. Hashimoto, S. Ohnuki
    日本金属学会・2011年秋期(第149回)講演大会  2011/11  沖縄コンベンションセンター
  • 中性子照射した改良SUS316鋼における組織と機械強度のミクロ−マクロ相関  [Not invited]
    山下 真一郎, 窪田 知宜, 岡 弘, 橋本 直幸, 大貫 惣明
    日本金属学会2011年秋期 (第149回)講演大会  2011/11  沖縄
  • Relationship between microstructure and mechanical properties in austenitic stainless steel  [Not invited]
    H.Oka, C. Liu, N.Hashimoto, S.Ohnuki, S.Yamashita
    15th International Conference on Fusion Reactor Materials (ICFRM-15)  2011/11  Charleston
  • Interface structure of oxide particle and its influence for helium in an ODS austenitic stainless steel  [Not invited]
    H.Oka, M.Watanabe, N.Hashimoto, S.Ohnuki, S.Yamashita, S.Ohtsuka
    15th International Conference on Fusion Reactor Materials (ICFRM-15)  2011/11  Charleston
  • ナノ組織から照射硬化まで  [Not invited]
    大貫 惣明, 岡 弘, 橋本 直幸
    東北大学金属材料研究所ワークショップ・「鉄鋼材料照射影響機構研究の最近の進展 〜ナノ組織から機械的特性へいかにつなげるか」  2011/09  仙台
  • オーステナイトステンレス鋼における微細組織と機械的特性の相関  [Not invited]
    岡 弘, 劉 伝歆, 橋本 直幸, 大貫 惣明, N. Su, S. Jiang, 山下 真一郎
    東北大学金属材料研究所ワークショップ・「鉄鋼材料照射影響機構研究の最近の進展 〜ナノ組織から機械的特性へいかにつなげるか」  2011/09  仙台
  • オーステナイト系ODS鋼の機械的特性および電子線照射下挙動  [Not invited]
    岡 弘, 橋本 直幸, 木下 博嗣, 柴山 環樹, 大貫 惣明
    日本金属学会・日本鉄鋼協会両北海道支部合同冬季講演大会  2011/01  室蘭工業大学
  • Characterization of oxide particles in ODS austenitic stainless steel after heavy ion irradiation up to high dose  [Not invited]
    H. Oka, Y. Yamazaki, H. Kinoshita, N. Hashimoto, S. Ohnuki, S. Yamashita, S. Ohtsuka
    2010 MRS fall meeting  2010/11  Boston
  • Stability of Y-Hf-O complex oxides in Fe-16Cr-4Al-0.6Hf ODS ferritic steel under electron-irradiation  [Not invited]
    C. Z. Yu, H. Oka, N. Hashimoto, S. Ohnuki
    Tenth Japan-China Symposium (JCS-10) on Materials for Advanced Energy Systems and Fission & Fusion Engineering  2010/10  Kyoto
  • Fe+イオン照射したオーステナイト系ODS鋼の酸化物粒子の安定性  [Not invited]
    山﨑 洋介, 岡 弘, Liu Chuanxin, 木下 博嗣, 橋本 直幸, 大貫 惣明, 鵜飼 重治
    日本金属学会2010年秋期 (第147回)講演大会  2010/09  札幌
  • Radiation-hardening and Microstructural Change in Ferritic Steels ion-irradiated by TIARA Facility to Low Doses  [Not invited]
    Liu Chuanxin, H. Oka, N. Hashimoto, S. Ohnuki, N. Okubo, I. Ioka, M. Ando, K. Shiba, S. Jitsukawa
    日本金属学会2010年秋期 (第147回)講演大会  2010/09  北海道大学
  • オーステナイト系ODS 鋼の酸化物粒子の形成挙動  [Not invited]
    岡 弘, 渡部 雅, 橋本 直幸, 柴山 環樹, 大貫 惣明, 山下 真一郎, 大塚 智史
    日本金属学会2010年秋期 (第147回)講演大会  2010/09  北海道大学
  • Multiple-beam Irradiation Effects in Reduced Activation Ferritic Steels  [Not invited]
    N. Hashimoto, H. Oka, H. Kinoshita, S. Ohnuki
    The 7th Pacific Rim International Conference on Advanced Materials and Processing (PRICM-7)  2010/08  Cairns
  • オーステナイト系ODS鋼の微細組織と機械的特性  [Not invited]
    岡 弘, 橋本 直幸, 木下 博嗣, 柴山 環樹, 大貫 惣明, 山下 真一郎, 大塚 智史
    日本金属学会2010年春期(第146回)講演大会  2010/03  筑波大学
  • Evaluation of micro-hardness and microstructure of ion-irradiated ODS steels  [Not invited]
    C. X. LIU, H. Oka, N. Hashimoto, S. Ohnuki, S. Yamashita, S. Ohtsuka, M. Ando, S. Jitsukawa
    日本鉄鋼協会・日本金属学会両北海道支部合同冬季講演大会  2010/01  北海道大学
  • Fabrication and mechanical properties of nano-particle dispersed austenitic stainless steels  [Not invited]
    M. Watanabe, H. Oka, N. Hashimoto, S. Ohnuki, T. Shibayama, S. Yamashita, S. Ohtsuka
    14th International Conference on Fusion Reactor Materials (ICFRM-14)  2009/09  Sapporo
  • Electron irradiation properties of ODS austenitic stainless steel modified by third element addition  [Not invited]
    H. Oka, M. Watanabe, N. Hashimoto, S. Ohnuki, T. Shibayama, H. Kinoshita, S. Yamashita, S. Ohtsuka
    14th International Conference on Fusion Reactor Materials (ICFRM-14)  2009/09  Sapporo
  • オーステナイト系ODS鋼の電子線照射欠陥形成に及ぼすヘリウムの影響  [Not invited]
    岡 弘, 渡部 雅, 橋本 直幸, 木下 博嗣, 柴山 環樹, 大貫 惣明
    日本鉄鋼協会・日本金属学会両北海道支部合同サマーセッション  2009/07  室蘭

Awards & Honors

  • 2018/09 日本原子力学会 材料部会 第10回 日本原子力学会 材料部会奨励賞
     9Cr-ODS鋼の微細組織と高温強度に及ぼす不純物窒素の影響 
    受賞者: 岡 弘
  • 2015/10 Forschungszentrum Jülich - ICFRM-17 ICFRM-17 Poster Award in Session 1
     Effect of thermo-mechanical treatment on nano-structure of 9Cr-ODS steel 
    受賞者: Hiroshi Oka

Research Grants & Projects

  • 高エネルギー粒子線照射下での損傷組織と強度の相関関係式の構築
    日本学術振興会:科学研究費助成事業 特別研究員奨励費
    Date (from‐to) : 2012 -2013 
    Author : 岡 弘
     
    1. 溶質原子の強度因子への影響 前年度において, 林転位の強度因子α_dが, 試験温度と固溶炭素量の影響を受けやすいことを指摘し, 本年度は溶体化処理したモデル合金を用いて上記を定量的に評価した. 一軸歪を導入し林転位の密度を変えたサンプルの降伏応力および転位密度から, α_dは固溶尿素量と試験温度に依存し, α_d (T, C)と表記できることを明らかにした. これは, 林転位として存在する拡張転位が炭素雰囲気に固着され, 炭素濃度が高い場合には転位の収縮に多くのエネルギーが必要なためである. 2. 熱時効材 熱時効したサンプルについて, 500℃時効では硬度は増加, 降伏応力は減少し, 700℃時効では硬度・降伏応力とも減少した. これは中性子照射材とほぼ同様の傾向であった. 700℃時効材での軟化は, 主に転位の回復によるものであり, 大きな析出物の強度への影響は小さいものと推察した. 3. 酸化物分散強化型316鋼 酸化物は平均径10nmほどで大半が陰イオン欠損型蛍石構造のY_2Hf_2O_7であり, その分散状態は不均一であった. 結晶粒径は平均1μmほどで, 通常の溶体化処理316鋼に比べ非常に小さく, 強度増加分のうち80%は結晶粒微細化によるものであった. これは, 酸化物の分散が不均一であるために有効粒子間距離が大きく, みかけの強度因子が小さくなったものであり(0.08-0.1程度), 均一分散の重要性を指摘した. 4. 妥当性検討と統合モデル 組織-強度相関についてはオロワンの式が有用であり, 強度因子αついては障害物の種類ごとに詳細に理解することが重要で, 特に林転位では温度と固溶炭素濃度に影響されることを明らかにした. 前年度結果を含む機械特性のスケール相関では, 基本的に線形関係として問題はないが, 一部特殊な場合(結晶粒が小さい場合, イオン照射材に代表されるような硬さの深さ依存が存在する場合, など)は適用できないことを示した.


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