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 C from 0.37 to 5.5510 n/m (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, 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.
Kondo, Keietsu; Miwa, Yukio*; Okubo, Nariaki; Kaji, Yoshiyuki; Tsukada, Takashi
Journal of Nuclear Materials, 417(1-3), p.892 - 895, 2011/10
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 330C to 550C. And it was revealed that aluminum addition to SSs was effective in the case of irradiation at elevated temperature. 316L/Al irradiated at 550C 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.
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
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 1.010m/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.
Sato, Tomonori; Miwa, Yukio; Tsukada, Takashi; Uchida, Shunsuke
Proceedings of 14th International Conference on Environmental degradation of Materials in Nuclear Power Systems (CD-ROM), p.1041 - 1052, 2009/08
At the surface of a pure metal, a particular half cell reaction corresponding to an exposed water chemistry condition occurs. In principle, O and HO concentrations under irradiation condition can be determined by the comparison of the redox potential responses of some kind of pure metals corresponding to exposed condition. We are developing a new array-type sensor based on this concept to determine O and HO concentrations in aqueous environment under irradiation. In order to confirm the concept of sensor, responses of the redox potentials of pure metals to the changes in O and HO concentrations were measured under non-irradiated condition. (1) The different responses of redox potentials to O and HO concentrations were obtained for the pure Fe, Cr, Zr and Pt electrodes. (2) The possibility of the concept of new array-type sensor to determine the O and HO concentrations was confirmed.
Nagashima, Nobuo*; Hayakawa, Masao*; Tsukada, Takashi; Kaji, Yoshiyuki; Miwa, Yukio*; Ando, Masami*; Nakata, Kiyotomo*
Atsuryoku Gijutsu, 47(4), p.236 - 244, 2009/07
In this study, micro-hardness tests and AFM observations were performed on SUS316L low-carbon austenitic stainless steel pre-strained by cold rolling to investigate its deformation behavior. The following results were obtained. Despite the fact that the same plastic strain was applied, post-tensile test AFM showed narrower slip-band spacing in a reduction in area of 30% cold-rolled specimen than the unrolled specimen. Concentrated slip bands were observed near grain boundaries. Micro-hardness exceeding 300 was found to occur frequently in after tensile test specimens with a reduction in area of 30% or more, particularly at grain boundaries. It is suggested that the nonuniformity of deformation at grain boundaries plays an important role of IGSCC crack propagation mechanism of low-carbon austenitic stainless steel.
Miwa, Yukio; Kaji, Yoshiyuki; Okubo, Nariaki; Kondo, Keietsu; Tsukada, Takashi
Nihon 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.
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.
Sato, Tomonori; Miwa, Yukio; Tsukada, Takashi; Uchida, Shunsuke
Zairyo To Kankyo 2009 Koenshu, p.63 - 66, 2009/05
At the surface of a pure metal, a particular half cell reaction corresponding to an exposed water chemistry condition occurs. In principle, O and HO concentrations in high temperature water can be determined by the comparison of the redox potential responses of some kind of pure metals corresponding to exposed condition. In order to confirm the concept of sensor, responses of the redox potentials of pure metals to the changes in O and HO concentrations were measured under non-irradiated condition and compared with the calculated potentials.
Izumo, Hironobu; Chimi, Yasuhiro; Ishida, Takuya; Kawamata, Kazuo; Inoue, Shuichi; Ide, Hiroshi; Saito, Takashi; Ise, Hideo; Miwa, Yukio; Ugachi, Hirokazu; et al.
JAEA-Technology 2009-011, 31 Pages, 2009/04
Regarding Irradiation Assisted Stress Corrosion Cracking (IASCC) for austenitic stainless steel of the light water reactor (LWR), a lot of data that concerns the post irradiation evaluation (PIE) is acquired. However, IASCC occurs in LWR condition. Therefore, it is necessary to confirm adequacy of the PIE data comparing the experiment data under the simulated LWR condition. Bigger specimen is needed to acquire the effective data for the destruction dynamics in the study of stress corrosion cracking under neutron irradiation condition. Therefore, development of a new crack growth unit which can load to bigger is necessary to the neutron irradiation test. As a result, a prospect was provided in the unit that could load to specimen by changing load mechanism to the lever type from the linear type. And also, in the development of crack propagation unit, some technical issues were extracted from the discussion of the unit structure adopting the LVDT (Linear Variable Differential Transformer).
Miwa, Yukio; Jitsukawa, Shiro; Tsukada, Takashi
Journal of Nuclear Materials, 386-388, p.703 - 707, 2009/04
In order to examine the stress corrosion cracking (SCC) susceptibility of reduced activation ferritic/martensitic steel, F82H, slow strain rate test (SSRT) was performed at various temperature in oxygenated or hydrogenated water. Test specimens of F82H were heat-treated at various temperature conditions, or were cold-worked to imitate radiation hardening and machined to make single edge notch, or were neutron-irradiated at 493 K to 3.4 dpa. It was found that in unirradiated specimen, IGSCC occurred when specimen was normalized only, and TGSCC occurred when cold-worked (over 23%) and notched specimen was tested by SSRT at 573 K in oxygenated water. In irradiated specimen, TGSCC occurred, when SSRT was conducted at 573 K in hydrogenated (DH = 1 ppm) water or when the notched specimen was tested by SSRT at 573 K in oxygenated (DO = 10 ppm) water.
Okubo, Nariaki; Miwa, Yukio; Kondo, Keietsu; Kaji, Yoshiyuki
Journal of Nuclear Materials, 386-388, p.290 - 293, 2009/04
no abstracts in English
Nakano, Junichi; Nemoto, Yoshiyuki; Miwa, Yukio; Usami, Koji; Tsukada, Takashi; Hide, Koichiro*
Journal of Nuclear Materials, 386-388, p.281 - 285, 2009/04
Crack initiation and crack growth processes of irradiation assisted stress corrosion cracking on stainless steels were studied by slow strain rate testing in oxygenated high temperature water at 561 K. In-situ observation was carried out during SSRT. Specimens of type 304 stainless steel were subjected to a solution annealing (SA), a thermally sensitization (TS), or a cold working (CW) and irradiated to 1.010 n/m (E 1 MeV) at 323 K in the Japan Material Testing Reactor (JMTR). Crack initiations were observed before the maximum stress would be reached for the CW material in in-situ observation. In fracture surface examination, the TS material exhibited almost intergranular stress corrosion cracking while mixtures of transgranular stress corrosion cracking and ductile dimple fracture were observed for the SA and the CW materials.
Maekawa, Masaki; Kawasuso, Atsuo; Hirade, Tetsuya; Miwa, Yukio
Materials Science Forum, 607, p.266 - 268, 2008/11
We have developed a positron microbeam using magnetic lenses based on the commercial scanning electron microscope. The minimum beam diameter was 1.9 micron on target. Two-dimensional image of S parameter was successfully obtained. Using this apparatus, S parameter distribution around the crack tip introduced by a stress corrosion cracking of a stainless steel was obtained. S parameter increases at the further region of the tip of the crack. This shows that vacancy type defects may be generated as crack precursor.
Tsukada, Takashi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Miwa, Yukio; Nakano, Junichi; Sato, Tomonori; Uchida, Shunsuke
Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10
Since the intergranular stress corrosion cracking (IGSCC) has been a major issue of degradation and failure of structural materials in the boiling water reactors (BWRs), various types of experiments were carried out to investigate IGSCC behavior under the environmental conditions simulated those in BWR. This paper describes a summary of the relevant experimental techniques and the experiences of two types of experiments performed by the authors. For the in-pile SCC experiments, IGSCC crack growth rates were obtained as a function of stress intensity factor in high temperature water. On the other hand, SCC and corrosion tests were performed on un-irradiated specimens in high temperature water by injecting hydrogen peroxide, HO to simulate water radiolysis condition. In order to understand IGSCC behavior under irradiation in the reactor core from a mechanistic viewpoint, combinations of various types of experiments are essentially required.
Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi; Hayakawa, Masao*; Nagashima, Nobuo*
JAEA-Research 2008-064, 118 Pages, 2008/08
This report describes a result of the research conducted by the Japan Atomic Energy Agency and the National Institute for Materials Science under contract with JNES that was concerned with a multi-scale analysis of plastic deformation behavior at the crack tip of SCC. In this research, analyses of the plastic deformation behavior and microstructure around the crack tip were performed in a nano-order scale. The hardness measured in nano, meso and macro scales was employed as a common index of the strength, and the essential data necessary to understand the SCC propagation behavior were acquired and analyzed that are mainly a size of plastic deformation region and a microstructural information in the region, e.g. data of crystallografy, microscopic deformation and dislocations at the inside of grains and grain boundaries.
Kato, Yoshiaki; Miwa, Yukio; Takada, Fumiki; Omi, Masao; Nakagawa, Tetsuya
JAEA-Testing 2008-005, 48 Pages, 2008/06
This report is concerned with the EBSD-OIM analyzer for irradiated reactor materials, which was installed in the JMTR hot laboratory. As the first time in the world, it was installed in a hot cell as one of the examination facilities for irradiated nuclear materials and contributes to studies on IASCC (irradiation aided stress corrosion cracking) and IGSCC (irradiation grain boundary stress corrosion cracking). Its maintenance and operating experiences were described.
Maekawa, Masaki; Kawasuso, Atsuo; Hirade, Tetsuya; Miwa, Yukio
Transactions of the Materials Research Society of Japan, 33(2), p.287 - 290, 2008/06
We present the development and application of a positron microprobe. Positrons from a small positron source developed by us were moderated by a solid neon moderator, and were focused onto the specimen using the optics of conventional scanning electron microscope (SEM). At present stage, the beam diameter was determined by scanning the beam across a knife-edge, and was found to be about 3.9 micrometer. Two dimensional scanning measurement of a cracked sample (stress-corrosion cracking) was performed. In the region distant from the tip of a crack, increase of S-parameter was found, which are possibly associated with the vacancy defects.
Ide, Hiroshi; Izumo, Hironobu; Ishida, Takuya; Saito, Takashi; Hanawa, Satoshi; Matsui, Yoshinori; Iwamatsu, Shigemi; Kanazawa, Yoshiharu; Miwa, Yukio; Kaji, Yoshiyuki; et al.
JAEA-Technology 2008-013, 32 Pages, 2008/03
Dissolved oxygen ions and chlorine ions concentration have been used as an evaluation index of stress corrosion cracking behavior for the light water reactor materials. In addition to these parameters, Electrochemical Corrosion Potential (ECP) was commonly used as the evaluation. Therefore, as a part of the IASCC irradiation tests, the irradiation test of the iron oxide type and the platinum type of ECP sensor were carried out under the BWR coolant condition. As a result, some measurements of ECP sensor succeed. However, it was clear that the improvement of ECP sensor is necessary. In this report, developed irradiation capsule ECP sensor is reported.
Igarashi, Takahiro; Miwa, Yukio; Kaji, Yoshiyuki; Tsukada, Takashi
Journal of Power and Energy Systems (Internet), 2(4), p.1188 - 1196, 2008/00
The two-dimensional intergranular stress corrosion cracking (IGSCC) growth model has been developed to simulate branching cracks of IGSCC. In the model, the IGSCC is grown using the "grain-scaled" factors such as the length and strength of grain boundary and so on. Especially, the corrosion of grain boundary and the influence of shear stress acting on the grain boundary are introduced in the model. Using the model, computer simulation of crack growth was carried out under several load conditions with changing the ratio of axial to shear stress against the grain boundary. As a result of the simulations, we found out that the cause of crack branching was the influence of shear stress against the grain boundary, and that the synergistic effect of shear stress and corrosion of grain boundary leads to the oblique crack growth.