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.
Chen, 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
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
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.
Murray, 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
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
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji
Journal of Nuclear Materials, 505, p.44 - 53, 2018/07
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 sodium bulk flow was set at 0.07 wt ppm in the calculation.
Nagasu, Ryosuke*; Tanabe, Daijiro*; Yokotsuka, Satoshi*; Kumazawa, Noriyuki*; Ajiki, Takaya*; Aizawa, Yusuke*; Naganawa, Hirochika; Nagano, Tetsushi; Yanase, Nobuyuki*; Mitamura, Hisayoshi*; et al.
Kankyo Joka Gijutsu, 17(2), p.58 - 61, 2018/03
A new technology to suppress cesium migration from forests has been developed collaboratively by Ibaraki University, Kumagai-gumi Co., Ltd. and its group company, Technos, and JAEA. The new technology utilizes polyelectrolytes (polymers with electric charges) and clay minerals to control Cs migration with the aid of natural forces such as rainfall and rainwater runoff. In Imitate-mura, Fukushima, verification tests of the new technology have been performed and its effect on controlling Cs migration from forests to grass farm adjoining the forests has been proven.
Chadwick, M. B.*; Capote, R.*; Trkov, A.*; Herman, M. W.*; Brown, D. A.*; Hale, G. M.*; Kahler, A. C.*; Talou, P.*; Plompen, A. J.*; Schillebeeckx, P.*; et al.
Nuclear Data Sheets, 148, p.189 - 213, 2018/02
The CIELO collaboration has studied neutron cross sections on nuclides that significantly impact criticality in nuclear facilities - U, U, Pu, Fe, O and H - with the aim of improving the accuracy of the data and resolving previous discrepancies in our understanding. This multi-laboratory pilot project, coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 40 with support also from the IAEA, has motivated experimental and theoretical work and led to suites of new evaluated libraries that accurately reflect measured data and also perform well in integral simulations of criticality. This report summarizes our results and outlines plans for the next phase of this collaboration.
Steppenbeck, D.*; Takeuchi, Satoshi*; Aoi, Nori*; Doornenbal, P.*; Matsushita, Masafumi*; Wang, H.*; Baba, Hidetada*; Go, Shintaro*; Holt, J. D.*; Lee, J.*; et al.
Physical Review C, 96(6), p.064310_1 - 064310_10, 2017/12
no abstracts in English
Tanno, Takashi; Takeuchi, Masayuki; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Materials, 494, p.219 - 226, 2017/10
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.
Shand, C. M.*; Podolyk, Zs.*; Grska, M.*; Doornenbal, P.*; Obertelli, A.*; Nowacki, F.*; Otsuka, T.*; Sieja, K.*; Tostevin, J. A.*; Tsunoda, T.*; et al.
Physics Letters B, 773, p.492 - 497, 2017/10
Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.
Asayama, Tai; Otsuka, Satoshi
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 15 Pages, 2017/06
This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be used for the fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr- and 11Cr-ODS steels have been extensively developed. Their laboratory-scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. As for the PNC-FMS wrapper tube, the development of a dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, codification of 316FR stainless steel and Modified 9Cr-1Mo steel is ongoing. Acquisition and collection of long-term data of base metal and welded joints are continued and evaluation methodologies are being developed to establish a technical basis for 60-year design.
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.
Journal of Nuclear Materials, 487, p.229 - 237, 2017/04
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 1400C 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 1200C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200C. This degradation was attributed to grain boundary sliding deformation with / 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 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).
Doornenbal, P.*; Scheit, H.*; Takeuchi, Satoshi*; Utsuno, Yutaka; Aoi, Nori*; Li, K.*; Matsushita, Masafumi*; Steppenbeck, D.*; Wang, H.*; Baba, Hidetada*; et al.
Physical Review C, 95(4), p.041301_1 - 041301_5, 2017/04
no abstracts in English
Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
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.
Ijiri, Yuta*; Ono, Naoko*; Ukai, Shigeharu*; Yu, H.*; Otsuka, Satoshi; Abe, Yosuke; Matsukawa, Yoshitaka*
Philosophical Magazine, 97(13), p.1047 - 1056, 2017/02
The interaction between oxide particles and dislocations in a 9Cr-ODS ferritic steel is investigated by both static and in-situ TEM observation under dynamic straining conditions and room temperature. The measured obstacle strength () of the oxide particles was no greater than 0.80 and the average was 0.63. The dislocation loops around some coarsened particles were also observed. The calculated obstacle strength by a stress formula of the Orowan interaction is nearly equaled to the average experimental value. Not only cross-slip system but also the Orowan interaction should be considered as the main interaction mechanism between oxide particles and dislocation in 9Cr-ODS ferritic steel.
Ukai, Shigeharu*; Otsuka, Satoshi; Kaito, Takeji; de Carlan, Y.*; Ribis, J.*; Malaplate, J.*
Structural Materials for Generation IV Nuclear Reactors, p.357 - 414, 2017/00
Oxide dispersion strengthened (ODS) steels are the most promising candidate materials for fuel cladding of generation IV nuclear reactors. The progress and current status for development of ODS/FM(ferrite-martensite) steels conducted mainly in Japan and France are overviewed. The chemical compositions of ODS/FM steels are listed. Fabrication routes of cladding tube are mentioned for ferrite-type ODS steels using recrystallized process and martensite-type one using - phase transformation. The optimized process is identical for both countries. Joining process between cladding and end-plug has been also developed by using the pressurized resistance upset welding method. The improvements brought by ODS/FM steels in high-temperature strength and irradiation resistance are verified.