Suzuki, Kenji*; Shobu, Takahisa
E-Journal of Advanced Maintenance (Internet), 10(4), p.9 - 17, 2019/02
In materials with an elastic anisotropy, a stress difference is generated between crystals when plastic deformation occurs, and it is known that this is deeply involved in material fracture. In this study, the residual stress for load direction in the plastically deformed material was investigated for each crystal orientation using the high-energy synchrotron radiation diffraction method. As a result, it was found that the residual stress is a tensile residual stress at an index with a high X-ray elastic constant (Young's modulus obtained for each diffraction surface) and a compressive residual stress at an index with a low X-ray elastic constant. We believe that this result will be useful for the technique of controlling the crystal orientation like the texture as improving the material strength.
Wan, T.; Saito, Shigeru
Metals, 8(8), p.627_1 - 627_22, 2018/08
Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00
In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at 288C on neutron-irradiated 316L stainless steels (SSs) at 12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at 2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.
Sekio, Yoshihiro; Yamagata, Ichiro; Akasaka, Naoaki; Sakaguchi, Norihito*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 8 Pages, 2017/06
The widths of void denuded zones (VDZs) which were formed near random grain boundaries by neutron irradiation were analyzed in order to perform quantitative evaluations for the irradiation-induced point defect behavior in the modified 316 stainless steel (PNC316) having been developed by JAEA. Namely, the temperature dependence of VDZ width was investigated and vacancy migration energy of the PNC316 steel was estimated from the VDZ width analysis for the neutron-irradiated specimens. The obtained value of vacancy migration energy was estimated as 1.46 eV, which was consistent with that from the exiting method using electron in-situ examination. This indicates that VDZ analysis could be effective method to evaluate especially vacancy migration energy during irradiation, and this would be realized from not in-situ observation but post-irradiation examination in the case of neutron irradiation.
Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka
Journal of Nuclear Materials, 480, p.386 - 392, 2016/11
This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.
Nemoto, Yoshiyuki; Oishi, Makoto; Ito, Masayasu; Kaji, Yoshiyuki
Nihon Hozen Gakkai Dai-12-Kai Gakujutsu Koenkai Yoshishu, p.105 - 112, 2015/07
Authors previously reported that Eddy current method and AC magnetization method have potential to be applied for development of diagnostic technics to detect the sign of material degradation before cracking on the austenitic stainless steels used as structural material in nuclear power plants. In typical austenitic stainless steels such as type304, magnetic ferrite phase would exist in the alloy before irradiation, and it is concerned to disturb the magnetic measurement on irradiated material. Magnetic measurements were conducted in this work on type304 austenitic stainless steel specimens irradiated up to different doses. In addition, microstructure observation was conducted on the area including grain boundary to discuss the correlation of magnetization on irradiated austenitic stainless alloy and grain boundary cracking. Obtained magnetic data on irradiated type304 stainless steel were seen clearly different from that on un-irradiated specimen, and showed positive correlation with radiation dose, therefore it was thought that magnetic measurement technics can be applied for the material which contains certain quantity of ferrite phase before irradiation. In the microstructural observation, magnetic phase (FeNi) formation along the grain boundary was revealed.
Tachibana, Yukio; Iyoku, Tatsuo
Nuclear Engineering and Design, 233(1-3), p.261 - 272, 2004/10
no abstracts in English
Kurata, Yuji; Futakawa, Masatoshi
Journal of Nuclear Materials, 325(2-3), p.217 - 222, 2004/09
The corrosion properties of three austenitic steels with different Si contents were studied under oxygen-saturated liquid Pb-Bi condition for 3000h. The three austenitic steels did not exhibit appreciable dissolution of Ni and Cr at 450C. At 550C, the thick ferrite layer produced by dissolution of Ni and Cr was found in JPCA and 316ss with low Si contents while the protective oxide film composed of Si and O was formed on 18Cr-20Ni-5Si steel and prevented dissolution of Ni and Cr. It was found that the Si added austenitic steel exhibited excellent corrosion resistance in liquid Pb-Bi.
Kikuchi, Kenji; Saito, Shigeru; Nishino, Yasuharu; Usami, Koji
Proceedings of 6th International Meeting on Nuclear Applications of Accelerator Technology (AccApp '03), p.874 - 880, 2004/00
Specimens irradiated at SINQ were tested by tensile and fatigue. Speciemns were irradiated by 580MeV proton beams under spallation reaction during two years, transported to JAERI and tested at JAERI Hot Cell. Material is JPCA austenitic stainless steel. Strain-to-necking is over 8% at 250C test temperature and are different from APT handbook database. Fatigue test was conducted at low stress regime of high cycle fatigue. The number of cycles to failure is reduced by factors five to ten. These data will help a design of spallation target in JPARC.
Zairyo To Kankyo, 52(2), p.66 - 72, 2003/02
Irradiation assisted stress corrosion cracking (IASCC) is a potential failure mode suffered by the core-components of austenitic stainless steels in the aged light-water reactor (LWR), which is the intergranular type cracking caused by synergistic effects of neutron/gamma radiation and chemical environment. Effects of radiation on the materials and high-temperature water are discussed in this paper to understand IASCC phenomenon from a mechanistic viewpoint. It is essential to elucidate the radiation-induced microcompositional and microstructural changes in the alloy for mechanistic and predictive investigations of IASCC. Although grain boundary segregations of alloying and impurity elements are significant factors affecting IASCC, it has been considered that the radiation-induced microstructural and mechanical changes of materials play critical roles in IASCC. For mechanistic understanding of IASCC, further fundamental research works with experimental and theoretical approaches are needed. Efforts directed to the researches at the Japan Atomic Energy Research Institute are also described.
Nemoto, Yoshiyuki; Miwa, Yukio; Tsuji, Hirokazu; Tsukada, Takashi
Dai-12-Kai MAGDA Konfarensu (Oita) Koen Rombunshu, p.191 - 196, 2003/00
Development and research about analytical method for the study of corrosion behavior of austenitic stainless steel after irradiation was conducted from the point of view for basic study of IASCC (Irradiation Assisted Stress Corrosion Cracking). Ion irradiations were conducted with several irradiation conditions these were irradiation temperature, radiation damage, the contents of helium (He) implantation. AFM (Atomic Force Microscope) was used to evaluate surface condition of irradiated specimens after corrosion procedure. Corrosion condition was developed to obtain good surface condition of irradiated specimens to evaluate corrosion behavior by AFM. It was succeeded and corrosion behavior at inside of grains and grain boundaries of irradiated specimens was obtained. EBSP (Electron Backscatter Diffraction Pattern) was used to evaluate relation of corrosion behavior with grain boundary character. Moreover, relations of corrosion behavior with irradiation condition were discussed.
Nakano, Junichi; Miwa, Yukio; Tsukada, Takashi; Kikuchi, Masahiko; Kita, Satoshi; Nemoto, Yoshiyuki; Tsuji, Hirokazu; Jitsukawa, Shiro
Journal of Nuclear Materials, 307-311(Part2), p.1568 - 1572, 2002/12
Type 316LN stainless steel of the international thermonuclear experimental reactor (ITER) Grade (316LN-IG SS) is being considered for the first wall/ blanket component. Hot isostatic pressing (HIP) technique is expected for the fabrication of module. To evaluate the integrity and susceptibility to stress corrosion cracking (SCC) of HIPed 316LN-IG SS, tensile tests in vacuum and slow strain rate tests (SSRT) in high temperature water were performed. Specimen with the HIPed joint shows no deterioration of the tensile strength and susceptibility to SCC in oxygenated water. Thermally sensitized specimen with the HIPed joint was low susceptible to SCC in creviced environment. It is concluded that the strength at joint location is as high as that at the base alloy and the joint interface appears integrity.
Kiuchi, Kiyoshi; Ioka, Ikuo; Tachibana, Katsumi; Suzuki, Tomio; Fukaya, Kiyoshi*; Inohara, Yasuto*; Kambara, Shozo; Kuroda, Yuji*; Miyamoto, Satoshi*; Ogura, Kazutomo*
JAERI-Research 2002-008, 63 Pages, 2002/03
no abstracts in English
Ioka, Ikuo; Miwa, Yukio; Tsuji, Hirokazu; Yonekawa, Minoru; Takada, Fumiki; Hoshiya, Taiji
JSME International Journal, Series A, 45(1), p.51 - 56, 2002/01
The low cycle creep-fatigue test with tensile strain hold of the austenitic stainless steel irradiated to 2dpa was carried out at 823K in vacuum. The applicability of creep-fatigue life prediction methods to the irradiated specimen was examined. The fatigue life on the irradiated specimen without tensile strain hold time was reduced by a factor of 2-5 in comparison with the unirradiated specimen. The fraction of intergranular fracture increased with increasing strain hold time. The decline in fatigue life of the irradiated specimen with tensile strain hold was almost equal to that of the unirradiated specimen. For the irradiated specimen, the time fraction damage rule trends to yield unsafe estimated lives and the ductility exhaustion damage rule trends to yield generous results. However, all of data were predicted within a factor of three on life by the linear damage rule.
Nemoto, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Kikuchi, Masahiko; Tsuji, Hirokazu
JAERI-Tech 2001-079, 25 Pages, 2001/12
Development and research about analytical method for the study of oxide film on austenitic stainless steel had been conducted from the point of view for basic study of IASCC (Irradiation Assisted Stress Corrosion Cracking). Nickel plating and copper plating had been compared as the oxide film protection while the fabrication for cross sectional observation. And thin film specimens for microstructural observation were fabricated using FIB (Focused Ion Beam) technique. Microstructure of oxide film on stainless steel had been observed with FE-TEM (Field Emission gun - Transmission Electron Microscope), and the chemical composition was analyzed with EDS (Energy dispersed X-ray Spectrometer). The oxide film had been formed in high pressure (8MPa) and high temperature (288) water, contains saturated oxygen. The thickness of oxide film was about 1m as maximum. Micro grains of Fe oxide with 100nm in diameter were formed in the oxide film. On the boundary with alloy, there was about 10nm thickness of passive film formed with Cr oxide.
Ioka, Ikuo; Yonekawa, Minoru; Miwa, Yukio; Mimura, Hideaki; Tsuji, Hirokazu; Hoshiya, Taiji
Journal of Nuclear Materials, 283-287(Part.1), p.440 - 445, 2000/12
no abstracts in English
Matsui, Yoshinori; Hoshiya, Taiji; Jitsukawa, Shiro; Tsukada, Takashi; Omi, Masao; ; Oyamada, Rokuro; *
Journal of Nuclear Materials, 233-237, p.188 - 191, 1996/00
no abstracts in English
Kato, Teruo; Maeta, Hiroshi
Teion Kogaku, 30(3), p.143 - 149, 1995/00
no abstracts in English
Purazuma, Kaku Yugo Gakkai-Shi, 70(7), p.719 - 725, 1994/07
no abstracts in English
Purazuma, Kaku Yugo Gakkai-Shi, 70(7), p.697 - 703, 1994/07
no abstracts in English