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Journal Articles

Hybridization of Bogoliubov quasiparticles between adjacent CuO$$_2$$ layers in the triple-layer cuprate Bi$$_2$$Sr$$_2$$Ca$$_2$$Cu$$_3$$O$$_{10+delta}$$ studied by angle-resolved photoemission spectroscopy

Ideta, Shinichiro*; Johnston, S.*; Yoshida, Teppei*; Tanaka, Kiyohisa*; Mori, Michiyasu; Anzai, Hiroaki*; Ino, Akihiro*; Arita, Masashi*; Namatame, Hirofumi*; Taniguchi, Masaki*; et al.

Physical Review Letters, 127(21), p.217004_1 - 217004_6, 2021/11

 Times Cited Count:1 Percentile:50.56(Physics, Multidisciplinary)

Journal Articles

Influence of microstructure on IASCC growth behavior of neutron irradiated type 304 austenitic stainless steels in simulated BWR condition

Kaji, Yoshiyuki; Miwa, Yukio*; Shibata, Akira; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*

International Journal of Nuclear Energy Science and Engineering, 2(3), p.65 - 71, 2012/09

Crack growth rate (CGR) tests have been conducted with neutron irradiated compact tension (CT) specimens. The specimens were irradiated in the core region of the Japan Materials Testing Reactor (JMTR) in simulated BWR water environments at 288 $$^{circ}$$C from 0.37 to 5.55$$times$$10$$^{25}$$ n/m$$^{2}$$ (E$$>$$ 1 MeV) (0.62 to 9.2 dpa). The CGRs of base metals in high electrochemical corrosion potential (ECP) condition with 10 $$<$$ stress intensity factor, K $$<$$ 30 MPam$$^{1/2}$$, increased with increasing neutron fluence until 2 dpa and the CGRs were almost the same from 2 to 10 dpa. We investigated the influence of microstructure on CGR by microstructure observation and local strain measurement around the precipitate. This paper will discuss the relationship between CGR and microstructure, radiation hardening, radiation induced segregation.

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

IASCC evaluation method for irradiated core internal structures in BWR power plants

Takakura, Kenichi*; Tanaka, Shigeaki*; Nakamura, Tomomi*; Chatani, Kazuhiro*; Kaji, Yoshiyuki

Proceedings of 2010 ASME Pressure Vessels and Piping Conference (PVP 2010) (CD-ROM), 10 Pages, 2010/07

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as a degradation of core internal components in light water nuclear reactor. Japan Nuclear Energy Safety organization (JNES) had been conducting a project related to IASCC as a part of safety research and development study for the aging management and maintenance of the nuclear power plants. Based on the JNES project results, JNES proposed "IASCC evaluation guide for BWR core internals". The purpose of this paper is to describe the background of the guide, especially crack growth rate (CGR) tests for irradiated stainless steels.

Journal Articles

Influence of microstructure on IASCC growth behavior of neutron irradiated type 304 austenitic stainless steels in simulated BWR condition

Kaji, Yoshiyuki; Miwa, Yukio; Shibata, Akira; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*

Proceedings of 14th International Conference on Environmental degradation of Materials in Nuclear Power Systems (CD-ROM), p.1181 - 1191, 2009/08

The CGR tests of neutron irradiated Type 304 SS were conducted in BWR conditions and the results were compared with those of Type 304L and 316L SS, and following results were obtained. (1) The CGR increase with increasing neutron fluence and the power law of K on the CGR was observed above F2 neutron fluence level (1.4 dpa). The different tendency is observed between Type 304 SS and L-grade SS (Type 304L and 316L SS) with increasing neutron fluence above F3 (4.3 dpa) level. (2) The CGR of Type 304 SS is slightly small as compared with those of Type 304L and 316L SS at the same neutron fluence and shows an increasing tendency above 4 dpa and reaches to 1.0$$times$$10$$^{-9}$$m/s in 9 dpa. (3) The neutron fluence dependence on uniform elongation is different with Type 304, 304L SS and Type 316L SS, that is, the neutron fluence in which the local deformation like channeling deformation is dominant, is high for Type 316L SS.

Journal Articles

IASCC crack growth rate of neutron irradiated low carbon austenitic stainless steels in simulated BWR condition

Chatani, Kazuhiro*; Takakura, Kenichi*; Ando, Masami*; Nakata, Kiyotomo*; Tanaka, Shigeaki*; Ishiyama, Yoshihide*; Hishida, Mamoru*; Kaji, Yoshiyuki

Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 9 Pages, 2007/00

Crack growth rate (CGR) tests have been conducted with neutron irradiated compact tension (CT) specimens. The CGR tests of 316L and 304L base metals irradiated from 0.516 to 1.07$$times$$10$$^{25}$$n/m$$^{2}$$ (E$$>$$1MeV), and of 316L and 308L weld metals irradiated from 0.523 to 0.541$$times$$10$$^{25}$$n/m$$^{2}$$ (E$$>$$1MeV) were performed using the reversing dc potential drop (DCPD) method under constant load at a few average stress intensity factors (K) and electrochemical corrosion potential (ECP) conditions at 288$$^{circ}$$C in water. CGRs of base metals were increased with increasing neutron fluence. Clear reductions in CGRs of base metals and weld metals were measured with decreasing ECP levels.

Journal Articles

Radiation damages of InGaAs photodiodes by high-temperature electron irradiation

Oyama, Hidenori*; Takakura, Kenichiro*; Nakabayashi, Masakazu*; Hirao, Toshio; Onoda, Shinobu; Kamiya, Tomihiro; Simoen, E.*; Claeys, C.*; Kuboyama, Satoshi*; Oka, Katsumi*; et al.

Nuclear Instruments and Methods in Physics Research B, 219-220, p.718 - 721, 2004/06

 Times Cited Count:3 Percentile:24.93(Instruments & Instrumentation)

no abstracts in English

Oral presentation

Investigation of SCC growth behavior with branching for irradiated materials

Kaji, Yoshiyuki; Igarashi, Takahiro; Miwa, Yukio; Taguchi, Taketoshi; Sozawa, Shizuo; Tsukada, Takashi; Hishida, Mamoru*; Takakura, Kenichi*

no journal, , 

no abstracts in English

Oral presentation

Study of the radiation damage effect in the commercial devices

Oyama, Hidenori*; Hayama, Kiyoteru*; Takakura, Kenichiro*; Hirao, Toshio; Onoda, Shinobu; Oshima, Takeshi; Kuboyama, Satoshi*; Arai, Manabu*

no journal, , 

no abstracts in English

Oral presentation

IASCC growth behavior evaluation of neutron-irradiated SUS304 stainless steel under BWR simulated high temperature water condition

Kaji, Yoshiyuki; Miwa, Yukio; Shibata, Akira; Kato, Yoshiaki; Taguchi, Taketoshi; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*

no journal, , 

SCC growth tests have been carried out using type 304 stainless steel that had been pre-irradiated 0.62 to 9.2dpa under BWR simulated high temperature water condition at 288$$^{circ}$$C in the JMTR. This paper describes the investigated results of crack growth rate characteristics from the point of view of microstructure, radiation hardening and radiation induced segregation.

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