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Journal Articles

Ultra-high temperature creep rupture and transient burst strength of ODS steel claddings

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

 Percentile:100(Materials Science, Multidisciplinary)

9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 $$^{circ}$$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 $$^{circ}$$C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 $$^{circ}$$C. This superior strength seemed to be owing to transformation of the matrix from the $$alpha$$-phase to the $$gamma$$-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.

Journal Articles

Effect of nitrogen concentration on nano-structure and high-temperature strength of 9Cr-ODS steel

Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji

Nuclear Materials and Energy (Internet), 16, p.230 - 237, 2018/08

Journal Articles

Model calculation of Cr dissolution behavior of ODS ferritic steel in high-temperature flowing sodium environment

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji

Journal of Nuclear Materials, 505, p.44 - 53, 2018/07

 Percentile:100(Materials Science, Multidisciplinary)

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.

Journal Articles

Corrosion behavior of ODS steels with several chromium contents in hot nitric acid solutions

Tanno, Takashi; Takeuchi, Masayuki; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Materials, 494, p.219 - 226, 2017/10

 Times Cited Count:4 Percentile:18.09(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 95$$^{circ}$$C. The corrosion rate exponentially decreased with effective solute chromium concentration (Cr$$_{eff}$$) 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$$_{eff}$$ 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$$_{eff}$$, concentrated nitric acid and addition of oxidizing ions helped to prevent the active dissolution, and suppressed the corrosion rate.

Journal Articles

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

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

 Times Cited Count:9 Percentile:4.24(Materials Science, Multidisciplinary)

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).

Journal Articles

Evaluation on tolerance to failure of ODS ferritic steel claddings at the accident conditions of fast reactors

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.

Journal Articles

Higher harmonic imaging of small defects in ODS steel cladding tubes and characterization of the defects with SEM

Kawashima, Koichiro*; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji

Dai-24-Kai Choompa Ni Yoru Hihakai Hyoka Shimpojiumu Koen Rombunshu (USB Flash Drive), p.99 - 104, 2017/01

no abstracts in English

Journal Articles

Application of FE-SEM to the measurement of U, Pu, Am in the irradiated MA-MOX fuel

Sasaki, Shinji; Tanno, Takashi; Maeda, Koji

Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 6 Pages, 2017/00

During irradiation in a fast reactor, the microstructure change of the mixed oxide fuels and the changes of element distributions occur because of a radial temperature gradient. Therefore, it is important to study the irradiation behavior of MA-MOX for advancement of fast reactor fuels. In order to make detailed observations of microstructure and elemental analyses of MA-MOX, irradiated MA-MOX specimens were carried out PIE by using a FE-SEM equipped with WDX. Because fuel samples have high radio activities and emit alpha-particles, the instrument was modified. the instrument was installed in a lead shield box and the control unit was separately located outside the box. The microstructure changes were observed in irradiated MA-MOX specimen. The characteristic X-rays peaks were detected successfully. By measuring the intensities of characteristic X-rays, it was tried quantitative analysis of U, Pu, Am along radial direction of irradiated specimen.

Journal Articles

Effect of thermo-mechanical treatments on nano-structure of 9Cr-ODS steel

Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Onuma, Masato*

Nuclear Materials and Energy (Internet), 9, p.346 - 352, 2016/12

 Times Cited Count:4 Percentile:31.55(Nuclear Science & Technology)

Journal Articles

Tensile properties and hardness of two types of 11Cr-ferritic/martensitic steel after aging up to 45,000 h

Yano, Yasuhide; Tanno, Takashi; Sekio, Yoshihiro; Oka, Hiroshi; Otsuka, Satoshi; Uwaba, Tomoyuki; Kaito, Takeji

Nuclear Materials and Energy (Internet), 9, p.324 - 330, 2016/12

BB2015-1728.pdf:1.04MB

 Times Cited Count:4 Percentile:31.55(Nuclear Science & Technology)

Journal Articles

Strength anisotropy of rolled 11Cr-ODS steel

Tanno, Takashi; Yano, Yasuhide; Oka, Hiroshi; Otsuka, Satoshi; Uwaba, Tomoyuki; Kaito, Takeji

Nuclear Materials and Energy (Internet), 9, p.353 - 359, 2016/12

BB2015-1727.pdf:6.74MB

 Times Cited Count:3 Percentile:42.29(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 $$^{circ}$$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.

Journal Articles

Weldability of dissimilar joint between PNC-FMS and Type 316 steel under electron beam welding

Yano, Yasuhide; Kaito, Takeji; Tanno, Takashi; Otsuka, Satoshi

Journal of Nuclear Science and Technology, 52(4), p.568 - 579, 2015/04

 Times Cited Count:1 Percentile:81.41(Nuclear Science & Technology)

The dissimilar butt welding joint of 11Cr-ferritic/martensitic steel (PNC-FMS) and Type 316 austenitic steel (SUS316) produced by electron beam (EB) welding was studied. This study was carried out to investigate optimization of EB welding and post weld heat treatment (PWHT). Optimum EB welding conditions were a focus position of 30-40 mm and a welding speed of 1750-2000 mm/min, and optimum PWHT was performed after welding at 690$$^{circ}$$C for 60 min. As a result, no formation of delta ferrite was observed adjacent to the fusion zone, and the mechanical properties of the welds were similar to those of the base material. In this regard, EB welding is a proper fusion welding process for dissimilar PNC-FMS and SUS316.

Journal Articles

Effects of manufacturing process on impact properties and microstructures of ODS steels

Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Tanaka, Kenya

Journal of Nuclear Materials, 455(1-3), p.480 - 485, 2014/12

 Times Cited Count:8 Percentile:25.95(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) steels are noticed as an advanced alloy durable to high-temperature and high-dose neutron irradiation environment. Japan Atomic Energy Agency, 9-12Cr-ODS martensite steels have been developed as the primary candidate material for fast reactor fuel cladding tube. They would be also good candidates for fusion reactor blanket material. In this work, two types of 11Cr-ODS steels were manufactured: pre-mix and full pre-alloy ODS steels. Tensile tests, creep tests, 1/3 sized Charpy impact tests and metallurgical observations were carried out on these steels. The impact properties of full pre-alloy ODS steel was shown to be much superior than that of pre-mix ODS steels. It was demonstrated that the full pre-alloy process noticeably improved the microstructure homogeneity (i.e. reduction of inclusions and pores). The ductility of full pre-alloy ODS steels were better than that of pre-mix ODS steels.

JAEA Reports

A Study on the method of an equivalent continuous body modelling using crack tensor theory in the Mizunami Underground Research Laboratory Project

Sanada, Hiroyuki; Sato, Toshinori; Tanno, Takeo*; Hikima, Ryoichi*; Tada, Hiroyuki*; Kumasaka, Hiroo*; Ishii, Takashi*; Sakurai, Hideyuki*

JAEA-Research 2014-006, 124 Pages, 2014/06

JAEA-Research-2014-006.pdf:11.26MB

Japan Atomic Energy Agency has been implementing the Mizunami Underground Research Laboratory (MIU) Project in order to develop the comprehensive investigation techniques for the geological environment and the engineering techniques to construct a deep underground laboratory in crystalline rock. In the rock mechanical study in the MIU Project, the development of the evaluation method for the excavation damaged zone due to excavation of shafts and research galleries is one of the important issues. In this report, crack tensor was calculated using the tunnel wall mapping and rock mechanical test results in the shaft and research galleries in the MIU. Two dimension excavation analysis was conducted at the Ventilation Shaft and GL -500 m Sub Stage using the calculated crack tensor at GL -500 m. Based on calculated crack tensor at GL 500 m, validation of the crack tensor at GL -500 m estimated during Phase I was verified. Relative error of crack tensor was calculated in order to examine variation of relative error to the scale of observation areas.

JAEA Reports

Mizunami Underground Research Laboratory Project annual report for fiscal year 2012

Hama, Katsuhiro; Mikake, Shinichiro; Nishio, Kazuhisa; Matsuoka, Toshiyuki; Ishibashi, Masayuki; Sasao, Eiji; Hikima, Ryoichi*; Tanno, Takeo*; Sanada, Hiroyuki; Onoe, Hironori; et al.

JAEA-Review 2013-050, 114 Pages, 2014/02

JAEA-Review-2013-050.pdf:19.95MB

Japan Atomic Energy Agency (JAEA) at Tono Geoscience Center (TGC) is pursuing a geoscientific research and development project namely the Mizunami Underground Research Laboratory (MIU) Project in crystalline rock environment in order to construct scientific and technological basis for geological disposal of High-level Radioactive Waste (HLW). The MIU Project has three overlapping phases: Surface-based Investigation phase (Phase I), Construction phase (Phase II), and Operation phase (Phase III). The MIU Project has been ongoing the Phase II and the Phase III in fiscal year 2012. This report presents the results of the investigations, construction and collaboration studies in fiscal year 2012, as a part of the Phase II and Phase III based on the MIU Master Plan updated in 2010.

JAEA Reports

Evaluation of irradiation behavior on oxide dispersion strengthened (ODS) steel claddings irradiated in Joyo/CMIR-6

Yano, Yasuhide; Otsuka, Satoshi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Sekine, Manabu; Endo, Toshiaki; Yamagata, Ichiro; Sekio, Yoshihiro; Tanno, Takashi; Uwaba, Tomoyuki; et al.

JAEA-Research 2013-030, 57 Pages, 2013/11

JAEA-Research-2013-030.pdf:48.2MB

It is necessary to develop the fast reactor core materials, which can achieve high-burnup operation improving safety and economical performance. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, the effects of fast neutron irradiation on mechanical properties and microstructure of 9Cr-and 12Cr-ODS steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the CMIR-6 at temperatures between 420 and 835$$^{circ}$$C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.

Journal Articles

Microstructure characterization of oxide dispersion strengthened steels containing metallic chromium inclusions after high-temperature thermal aging

Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji; Tanaka, Kenya

Materials Transactions, 54(10), p.2018 - 2026, 2013/10

 Times Cited Count:3 Percentile:67.03(Materials Science, Multidisciplinary)

Microstructure characterizations of 9Cr-oxide dispersion strengthened (ODS) steels were carried out after high-temperature thermal aging to reproduce the anomalous microstructure change that occurred in the BOR-60 irradiation test-formation of abnormally coarse and irregular precipitates a few tens of micrometers in size. In the 9Cr-ODS steel containing metallic Cr inclusions, coarse and irregular precipitates were formed nearby metallic Cr inclusions after the 750$$^{circ}$$C thermal aging for 8,000h. Based on the analyses using energy dispersive X-ray spectrometry (EDX) and electron backscattered pattern (EBSP), coarse and irregular precipitates were identified as M23C6.

JAEA Reports

Mizunami Underground Research Laboratory Project annual report for fiscal year 2011

Kunimaru, Takanori; Mikake, Shinichiro; Nishio, Kazuhisa; Tsuruta, Tadahiko; Matsuoka, Toshiyuki; Ishibashi, Masayuki; Sasao, Eiji; Hikima, Ryoichi; Tanno, Takeo; Sanada, Hiroyuki; et al.

JAEA-Review 2013-018, 169 Pages, 2013/09

JAEA-Review-2013-018.pdf:15.71MB

Japan Atomic Energy Agency (JAEA) at Tono Geoscience Center (TGC) is pursuing a geoscientific research and development project namely the Mizunami Underground Research Laboratory (MIU) Project in crystalline rock environment in order to construct scientific and technological basis for geological disposal of High-level Radioactive Waste (HLW). The MIU Project has three overlapping phases: Surface-based Investigation phase (Phase I), Construction phase (Phase II), and Operation phase (Phase III). The MIU Project has been ongoing the Phase II and the Phase III in 2011 fiscal year. This report shows the results of the investigation, construction and collaboration studies in fiscal year 2011, as a part of the Phase II and Phase III based on the MIU Master Plan updated in 2010.

Journal Articles

Microstructure and high-temperature strength of high Cr ODS tempered martensitic steels

Otsuka, Satoshi; Kaito, Takeji; Tanno, Takashi; Yano, Yasuhide; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Materials, 442(1-3), p.S89 - S94, 2013/09

 Times Cited Count:9 Percentile:25.82(Materials Science, Multidisciplinary)

The manufacturing tests of 11-12Cr ODS tempered martensitic steels were carried out, and their ferritic/martensitic duplex structures were quantitatively evaluated by three types of methods, i.e. high temperature XRD, EPMA and metallography. It was demonstrated that excessive formation of residual-alpha ferrite provided by increasing Cr can be suppressed by appropriately controlling the concentration of ferrite-forming element and austenite-forming element on the basis of the parameter "chemical driving force of $$alpha$$ to $$gamma$$ reverse transformation" as a useful indication. The 11Cr-ODS steel containing a small portion of residual-alpha ferrite was successfully manufactured. In the as-received condition, this 11Cr-ODS steel is shown to have the satisfactory creep strength and ductility as high as the 9Cr-ODS steel while 0.2% proof strength at 973K is lower than in the 9Cr-ODS steel.

Journal Articles

Evaluation of mechanical properties and nano-meso structures of 9-11%Cr ODS steels

Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Oba, Yojiro*; Onuma, Masato*; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Materials, 440(1-3), p.568 - 574, 2013/09

 Times Cited Count:7 Percentile:34.56(Materials Science, Multidisciplinary)

This study carried out mechanical tests and microstructure characterizations of several 9Cr and 11Cr-ODS tempered martensitic steels, and discussed the appropriate chemical composition range of 11Cr-ODS tempered martensitic steel from the viewpoint of high-temperature strength improvement. It was shown that the residual $$alpha$$-ferrite fraction in 11Cr-ODS steel was successfully controlled to the same level as the 9Cr-ODS steel by selecting the matrix chemical compositions on the basis of the multi-component phase diagram. The tensile strength decreased with decreasing W content from 2.0 to 1.4 wt%. On the other hand, creep strength at 973 K did not degrade by the decreasing W content. Both tensile strength and creep strength increased with increasing population of the nano-sized oxide particles. Small angle X-ray scattering analysis revealed that titanium and excess oxygen contents were key parameters in order to improve the dispersion condition of nano-sized oxide particles.

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