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Journal Articles

High-temperature short-range order in Mn$$_3$$RhSi

Yamauchi, Hiroki; Sari, D. P.*; Watanabe, Isao*; Yasui, Yukio*; Chang, L.-J.*; Kondo, Keietsu; Ito, Takashi; Ishikado, Motoyuki*; Hagiwara, Masato*; Frontzek, M. D.*; et al.

Communications Materials (Internet), 1, p.43_1 - 43_6, 2020/07

High-temperature short-range order is discovered up to 720 K in Mn$$_3$$RhSi by complementary use of neutron scattering and muon spin relaxation measurements.

Journal Articles

FE-SEM observation of chains of nanohillocks in SrTiO$$_{3}$$ and Nb-doped SrTiO$$_{3}$$ surfaces irradiated with swift heavy ions

Kitamura, Akane; Ishikawa, Norito; Kondo, Keietsu; Yamamoto, Shunya*; Yamaki, Tetsuya*

Nuclear Instruments and Methods in Physics Research B, 460, p.175 - 179, 2019/12

 Times Cited Count:0 Percentile:100(Instruments & Instrumentation)

Irradiation at grazing incidence formed chains of multiple hillocks on the surface of strontium titanate (SrTiO$$_{3}$$) and titanium oxide (TiO$$_{2}$$). They were observed with an atomic force microscope (AFM), however, the AFM measurement gives resolution errors in a nanometer order due to the curvature of the probe tip. To prevent these errors, a field emission scanning electron microscope (FE-SEM) would be a better option for observation. In this study, we performed SEM observations for the chains of the multiple hillocks. Single crystals of SrTiO$$_{3}$$ and TiO$$_{2}$$ were irradiated with 200 MeV $$^{136}$$Xe$$^{14+}$$ in the tandem accelerator at JAEA-Tokai. It was revealed that a lot of isolated hillocks were formed in a line on these surface. The diameter and the interval of those hillocks are discussed in comparison to AFM observation.

Journal Articles

FE-SEM observations of multiple nanohillocks on SrTiO$$_{3}$$ irradiated with swift heavy ions

Kitamura, Akane; Ishikawa, Norito; Kondo, Keietsu; Fujimura, Yuki; Yamamoto, Shunya*; Yamaki, Tetsuya*

Transactions of the Materials Research Society of Japan, 44(3), p.85 - 88, 2019/06

Swift heavy ions can create nanosized hillocks on the surfaces of various ceramics. When these materials are irradiated with swift heavy ions at normal incidence, each ion impact results in the formation of a single hillock on the surfaces. In contrast, irradiation at grazing incidence forms chains of multiple hillocks on the surface, for example, for strontium titanate (SrTiO$$_{3}$$). So far, chains of multiple hillocks have been investigated using atomic force microscopy (AFM). It should be noted that AFM measurements involve systematic errors of several nanometers due to the finite size of the probe tip. Consequently, it is possible that the image of one hillock may merge with that of a neighboring hillock even if the two hillocks are well separated. In contrast to AFM, field-emission scanning electron microscopy (FE-SEM) is a useful technique for obtaining higher-resolution images. In this study, we observed multiple nanohillocks on the surfaces of SrTiO$$_{3}$$ using FE-SEM. Crystals of SrTiO$$_{3}$$(100) and 0.05 wt% Nb-doped SrTiO$$_{3}$$(100) were irradiated with 350 MeV Xe ions, respectively, at grazing incidence, where the angle between the sample surface and the beam was less than 2$$^{circ}$$. On the SrTiO$$_{3}$$ surface, a chain of periodic nanohillocks is created along the ion path. In contrast, black lines accompanied by hillocks are observed on the Nb-doped SrTiO$$_{3}$$ surface.

Journal Articles

Structure of nitride layer formed on titanium alloy surface by N$$_{2}$$-gas exposure at high temperatures

Takeda, Yusuke; Iida, Kiyoshi*; Sato, Shinji*; Matsuo, Tadatoshi*; Nagashima, Yasuyuki*; Okubo, Nariaki; Kondo, Keietsu; Hirade, Tetsuya

JPS Conference Proceedings (Internet), 25, p.011023_1 - 011023_3, 2019/03

In this study, we prepared samples under two different conditions, (1) 810$$^{circ}$$C, for 600 min, and (2) 850$$^{circ}$$C, for 720 min. A depth-profile analysis of the surfaces of the samples is conducted through Doppler broadening (DB) measurements of positron annihilation $$gamma$$ rays using a slow positron beam. It was indicated that many of positrons annihilated in defects near the surface. According to the TEM image, there are nano-crystal grains near the surface and then positrons can diffuse in the grains and annihilate in defects at the grain boundaries. Furthermore, DB measurements indicated that there is a depth dependence on the chemical composition where positrons annihilate. EDS spectroscopy measurements also indicated that there is a depth dependence of impurities such as Vanadium. These results indicated change of the chemical composition at the grain boundaries.

Journal Articles

Ion irradiation effects on FeCrAl-ODS ferritic steel

Kondo, Keietsu; Aoki, So; Yamashita, Shinichiro; Ukai, Shigeharu*; Sakamoto, Kan*; Hirai, Mutsumi*; Kimura, Akihiko*

Nuclear Materials and Energy (Internet), 15, p.13 - 16, 2018/05

 Times Cited Count:4 Percentile:26.29(Nuclear Science & Technology)

Radiation hardening and microstructural evolution of ion irradiated 12Cr-6Al ODS ferritic steel was studied. Ion irradiation experiments were performed using Fe ions up to the nominal displacement damage of 20 dpa at the irradiation temperature was 300$$^{circ}$$C. The monotonical increase of radiation hardening with dose was confirmed by experimentally obtained hardness data. The radiation hardening was also calculated theoretically by introducing the microstructural character examined by TEM into the dispersed barrier hardening model. The results showed a good agreement with the experimentally obtained data up to 5 dpa, while a slight discrepancy was found between theoretical and experimental hardness values at 20 dpa. Radiation hardening was mainly caused by irradiation-induced defect clusters below the irradiation dose of 5 dpa. As the irradiation dose increased toward 20 dpa, an additional influence of the radiation appeared, which was assumed to be induced by $$alpha$$' phase transformation.

Journal Articles

Effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels

Aoki, So; Kondo, Keietsu; Kaji, Yoshiyuki; Yamamoto, Masahiro

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.663 - 672, 2018/00

 Times Cited Count:0 Percentile:100

The objective of this study was to clarify effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels. Specimens used were Type 304L and 316L steels. Both steels were cold rolled to 20% thickness reduction (CW) and then followed by long-term thermal aging at 288$$^{circ}$$C for 14,000 h (LTA). Crevice Bent Beam (CBB) test was carried out to estimate the SCC initiation susceptibility under BWR simulated water condition at high temperature. The present results of the CBB tests showed that 304L CW + LTA exhibited no SCC susceptibility. In contrast, the SCC initiation susceptibility of 316L increased by the combination of cold rolling and long-term thermal aging. To understand these results, evaluation on the changes in microchemistry, microstructure and mechanical properties induced by the CW and LTA treatment has been developed, and their correlation with the SCC initiation susceptibility is discussed.

Journal Articles

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu; Nakashima, Kazuo*; Kanazawa, Toru*; Tojo, Masayuki*

Journal of Nuclear Materials, 488, p.22 - 32, 2017/05

AA2016-0383.pdf:0.86MB

 Times Cited Count:1 Percentile:79.18(Materials Science, Multidisciplinary)

The authors previously conducted thermogravimetric analyses on zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study.

Journal Articles

Structure of nitride layer formed on titanium alloy surface by N$$_{2}$$-gas exposure at high temperatures

Takeda, Yusuke; Iida, Kiyoshi*; Sato, Shinji*; Matsuo, Tadatoshi*; Nagashima, Yasuyuki*; Okubo, Nariaki; Kondo, Keietsu; Hirade, Tetsuya

Journal of Physics; Conference Series, 791(1), p.012022_1 - 012022_4, 2017/02

 Times Cited Count:0 Percentile:100

Titanium alloy is widely used for applications such as golf club heads and structural materials for aircrafts. The surface can be exceedingly hardened by nitriding treatment that initiates defects, but there are some difficulties on use of titanium nitride because the layer can be exfoliated by stress. Therefore, we prepared samples in two different treatment conditions, (1) 810$$^{circ}$$C 600 min and (2) 850$$^{circ}$$C 720 min and performed depth profile analysis of Doppler broadening of positron annihilation $$gamma$$-rays (DB) for these samples. According to a calculation of nitrogen diffusion depth, the nitride layer should be only about 0.05-0.1$$mu$$m. However, the depth profile analysis of the DB measurement indicated that the defects introduced by nitriding treatment extended to a depth of 0.5$$mu$$m.

Journal Articles

Silver photo-diffusion and photo-induced macroscopic surface deformation of Ge$$_{33}$$S$$_{67}$$/Ag/Si substrate

Sakaguchi, Yoshifumi*; Asaoka, Hidehito; Uozumi, Yuki; Kondo, Keietsu; Yamazaki, Dai; Soyama, Kazuhiko; Ailavajhala, M.*; Mitkova, M.*

Journal of Applied Physics, 120(5), p.055103_1 - 055103_10, 2016/08

 Times Cited Count:8 Percentile:46.15(Physics, Applied)

Journal Articles

Development of corrosion-resistant improved Al-doped austenitic stainless steel

Kondo, Keietsu; Miwa, Yukio*; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi

Journal of Nuclear Materials, 417(1-3), p.892 - 895, 2011/10

 Times Cited Count:3 Percentile:69.1(Materials Science, Multidisciplinary)

For the purpose to suppress the degradation of corrosion resistance induced by irradiation in austenitic stainless steels (SSs), aluminum-doped type 316L SS (316L/Al) was fabricated, and its electrochemical corrosion property was estimated after Ni-ion irradiation at the temperature range from 330$$^{circ}$$C to 550$$^{circ}$$C. And it was revealed that aluminum addition to SSs was effective in the case of irradiation at elevated temperature. 316L/Al irradiated at 550$$^{circ}$$C up to 12 dpa showed high corrosion resistance in the vicinity of grain boundaries (GBs) and in grains, while the severe GB etching and local corrosion in grains were observed in irradiated 316L and 316 SS. It is supposed that the aluminum enrichment, which is caused by radiation induced segregation at GBs and by radiation induced precipitation such as Ni3Al in grains, was enhanced by high-temperature irradiation, and contributes to compensate the lost corrosion resistance by the chromium depletion.

Journal Articles

Stress corrosion cracking behavior of type 304 stainless steel irradiated under different neutron dose rates at JMTR

Kaji, Yoshiyuki; Kondo, Keietsu; Aoyagi, Yoshiteru; Kato, Yoshiaki; Taguchi, Taketoshi; Takada, Fumiki; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Takakura, Kenichi*; et al.

Proceedings of 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (CD-ROM), p.1203 - 1216, 2011/08

In order to investigate the effect of neutron dose rate on tensile property and irradiation assisted stress corrosion cracking (IASCC) growth behavior, the crack growth rate (CGR) test, tensile test and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating the temperature of boiling water reactor (BWR) up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). The radiation hardening increased with the dose rate, but there was little effect on CGR. Increase of the yield strength of specimens irradiated with the low dose rate condition was caused by the increase of number density of frank loops. Little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. Furthermore, there was little effect on local plastic deformation behavior near crack tip in the crystal plasticity simulation.

Journal Articles

Interrelationship between true stress-true strain behavior and deformation microstructure in the plastic deformation of neutron-irradiated or work-hardened austenitic stainless steel

Kondo, Keietsu; Miwa, Yukio; Tsukada, Takashi; Yamashita, Shinichiro; Nishinoiri, Kenji

Journal of ASTM International (Internet), 7(1), p.220 - 237, 2010/01

no abstracts in English

Journal Articles

New evaluation method of material degradation considering synergistic effects of radiation damage

Miwa, Yukio; Kaji, Yoshiyuki; Okubo, Nariaki; Kondo, Keietsu; Tsukada, Takashi

Nippon Kikai Gakkai M&M 2007 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), p.236 - 237, 2009/07

In core structural materials of next generation reactors, materials' degradation behavior by neutron irradiation damage and thermal (cyclic) stress should be considered with fair accuracy in design process, because the materials are used under higher temperature gradients and higher neutron flux fields than those in the present light water reactors. In the current experiential design rules, service lives of core structural components were determined by the materials degradation such as the increase of ductile-to-brittle transition temperature after post irradiation examination data. However, other materials degradations such as irradiation-assisted stress corrosion cracking (IASCC), which occurs by the degradation synergistically interacting with radiation hardening, local chemical composition change, swelling and radiation creep, should be considered reasonably in the design process of the next generation reactors, because of the anticipation of the beneficial effects by synergy of radiation damage. To predict material failure by IASCC with reasonable accuracy, in this study, each material degradation phenomenon with different dose dependence was modeled with consideration of radiation induced stress relaxation. In this paper, the models obtained by ion-irradiation experiments and compared by data from neutron irradiation experiments were presented, and the concept of our new evaluation method and the programming code for the failure simulation were outlined.

Journal Articles

New concept of damage evaluation method for core internal materials considering radiation induced stress relaxation, 2; Simulation of material degradation behavior using integrated model

Kaji, Yoshiyuki; Miwa, Yukio; Kondo, Keietsu; Okubo, Nariaki; Tsukada, Takashi

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), P. 9359, 2009/05

In this paper, we describe the simulation results of the irradiation assisted stress corrosion cracking (IASCC) behavior at the flaws considering the radiation induced stress relaxation (RISR) with residual stress introduced by the welding process for a long operation period.

Journal Articles

Effects of residual stress on irradiation hardening in stainless steels

Okubo, Nariaki; Miwa, Yukio; Kondo, Keietsu; Kaji, Yoshiyuki

Journal of Nuclear Materials, 386-388, p.290 - 293, 2009/04

 Times Cited Count:4 Percentile:65.47(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Influence of laser irradiation condition on a femtosecond laser-assisted tomographic atom probe

Nishimura, Akihiko; Nogiwa, Kimihiro; Otobe, Tomohito; Okubo, Tadakatsu*; Hono, Kazuhiro*; Kondo, Keietsu; Yokoyama, Atsushi

Ultramicroscopy, 109(5), p.467 - 471, 2009/04

 Times Cited Count:6 Percentile:69.95(Microscopy)

Influence of femtosecond laser pulse condition on the performance of an energy compensated tomographic atom probe (ECOTAP) was investigated. Chirping ratio for laser pulses was controlled by a compressor stage. We have succeeded to get tomographic images of oxide dispersion strengthen steel, which will be used as fast breeder reactors. The ECOTAP successfully observed that the instability of the femtosecond laser pulses make the mass peaks slightly sifted or broadened to higher mass number. To investigate insulator materials, numerical simulation of conductivity increase on diamond has been successfully demonstrated.

Journal Articles

New evaluation method of material degradation considering synergistic effects of radiation damage

Miwa, Yukio; Kaji, Yoshiyuki; Okubo, Nariaki; Kondo, Keietsu; Tsukada, Takashi

Journal of Solid Mechanics and Materials Engineering (Internet), 2(1), p.145 - 155, 2008/00

In core structural materials of next generation reactors, materials' degradation behavior by neutron irradiation damage and thermal (cyclic) stress should be considered with fair accuracy in design process (including maintenance and repair plans), because the materials are used under higher temperature gradients and higher neutron flux fields than those in the present light water reactors. In the current experiential design rules, service lives of core structural components were determined by the materials degradation such as the increase of ductile-to-brittle transition temperature after post irradiation examination data. However, other materials degradations such as irradiation-assisted stress corrosion cracking (IASCC) should be considered reasonably in the design process of the next generation reactors, because of the anticipation of the beneficial effects by synergistics of these radiation damage such as radiation hardening, local chemical composition change, swelling and radiation creep. To predict material failure by IASCC with reasonable accuracy, in this study, each material degradation phenomenon with different dose dependence was modeled with consideration of radiation induced stress relaxation. The models were integrated to simulate the failure behavior for the reactor operation period. In this paper, the models obtained by ion-irradiation experiments were presented, and the concept of new evaluation method and the programming code for the failure simulation were outlined.

Journal Articles

The Effects of residual stress on corrosion behavior of ion irradiated type 316L stainless steel

Kondo, Keietsu; Miwa, Yukio; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi

Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 11 Pages, 2007/00

The effect of residual stress on corrosion behavior in type 316L austenitic stainless steel was examined by ion irradiated specimens. Ion irradiation was performed on specimens both undeformed and deformed plastically by bending constrait at 330$$^{circ}$$C to average displacement damage from 1 to 45dpa. It was observed in EPR testing that deformed specimens showed higher corrosion resistance than undeformed specimens. Three-dimensional atom probe analysis was conducted on irradiated specimens. It was found that the enrichment of Ni, Si and the depletion of Cr at dislocations, and the degree of segregation was greater in undeformed specimen than in deformed specimen. It could be suggested that radiation induced segregation behavior of solute atoms as a consequence of diffusion and annihilation of irradiation defects at sink is affected by residual stress, and this also might affect the corrosion resistance.

JAEA Reports

Analyses of core Shroud materials by three dimensional atom probe (Contract research)

Kondo, Keietsu; Nemoto, Yoshiyuki; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Nagai, Yasuyoshi*; Hasegawa, Masayuki*; Okubo, Tadakatsu*; Hono, Kazuhiro*

JAEA-Research 2006-013, 39 Pages, 2006/12

JAEA-Research-2006-013.pdf:4.57MB

There has been an increasing number of stress corrosion cracking (SCC) incidents on low carbon austenitic stainless steels used in boiling water reactor (BWR) environments. To reveal the acceleration factor of intergranular crack propagation from the viewpoint of solute distribution in stainless steels, the material extracted from a core shroud of Japanese BWR was analyzed by the three dimensional atom probe (3DAP), which has the highest spatial resolution among the various microanalytical techniques. It was revealed by statistical analysis on 3DAP data that solute elements, such as Fe, Cr, Ni, Mo, Mn, Si, are randomly distributed in matrix of the shroud material. This result means that solute was not segregated or precipitated and was not form spinodal decomposition during the service. The concentration profile in the vicinity of grain boundary obtained from 3DAP dataset showed the random distribution of Cr. This result shows that degradation of the corrosion resistance induced by depletion of Cr was not responsible for the crack propagation along grain boundaries in low carbon stainless steel. On the other hand, enrichment of Mo and Si was observed at grain boundary. The width of the enriched zone was about 2 nm across the grain boundary, and the concentration of those elements could be much higher than the concentration obtained by field emission transmission electron microscopy/energy dispersive X-ray spectroscopy (FE-TEM/EDS). Therefore, it is necessary to study about the effects of enrichment of Mo and Si as a potential contributor to SCC.

Journal Articles

PIE technologies for the study of stress corrosion cracking of reactor structural materials

Ugachi, Hirokazu; Nakano, Junichi; Nemoto, Yoshiyuki; Kondo, Keietsu; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi; Kizaki, Minoru; Omi, Masao; Shimizu, Michio

JAEA-Conf 2006-003, p.253 - 265, 2006/05

Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in the light water reactors (LWRs) for a long period. In general, IASCC can be reproduced on the materials irradiated over a certain threshold fluence level of fast neutron by the post-irradiation examinations (PIEs) at hot laboratories. On the other hand, recently in the Japanese boiling water reactor (BWR) power plants, many incidents of stress corrosion cracking (SCC) of structural material such as the reactor core shrouds and primary loop recirculation (PLR) system piping were reported. In order to investigate the cause of SCC, PIEs at hot laboratories were carried out on the sample material extracted from BWR power plants. SCC studies require various kind of PIE techniques, because the SCC is caused by a complicated synergistic effects of stress and chemical environment on material that suffered degradations by irradiation and/or thermal aging. In this paper, we describe the PIE techniques adopted recently for our SCC studies, especially the crack growth measurement, uniaxial constant load (UCL) tensile test method, in-situ observation during slow strain rate test (SSRT) and several metallurgical test techniques using the FEtype transmission electron microscopy (FE-TEM), focused ion beam (FIB) processing technique, three Dimensional Atom Probe (3DAP) analysis and atomic force microscopy (AFM).

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