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Ikeuchi, Hirotomo; Sasaki, Shinji; Onishi, Takashi; Nakayoshi, Akira; Arai, Yoichi; Sato, Takumi; Ohgi, Hiroshi; Sekio, Yoshihiro; Yamaguchi, Yukako; Morishita, Kazuki; et al.
JAEA-Data/Code 2023-005, 418 Pages, 2023/12
JAEA-Data-Code-2023-005-01.pdf:24.59MBJAEA-Data-Code-2023-005-02.pdf:32.18MB
For safe and steady decommissioning of Tokyo Electric Power Company Holdings' Fukushima Daiichi Nuclear Power Station (1F), information concerning composition and physical/chemical properties of fuel debris generated in the reactors should be estimated and provided to other projects conducting the decommissioning work including the retrieval of fuel debris and the subsequent storage. For this purpose, in FY2021, samples of contaminants (the wiped smear samples and the deposits) obtained through the internal investigation of the 1F Unit 2 were analyzed to clarify the components and to characterize the micro-particles containing uranium originated from fuel (U-bearing particles) in detail. This report summarized the results of analyses performed in FY2021, including the microscopic analysis by SEM and TEM, radiation analysis, and elemental analysis by ICP-MS, as a database for evaluating the main features of each sample and the probable formation mechanism of the U-bearing particles.
and O in a reaction cell in triple quadrupole inductively coupled plasma mass spectrometryKazama, Hiroyuki; Konashi, Kenji*; Suzuki, Tatsuya*; Koyama, Shinichi; Maeda, Koji; Sekio, Yoshihiro; Onishi, Takashi; Abe, Chikage*; Shikamori, Yasuyuki*; Nagai, Yasuyoshi*
Journal of Analytical Atomic Spectrometry, 38(8), p.1676 - 1681, 2023/07
Times Cited Count：3 Percentile：56.33(Chemistry, Analytical)Sekio, Yoshihiro; Sakaguchi, Norihito*
Materials Transactions, 60(5), p.678 - 687, 2019/05
Times Cited Count：6 Percentile：30.55(Materials Science, Multidisciplinary)The quantitative evaluation of vacancy migration energies in high nickel model alloy was conducted by analyzing the void denuded zone (VDZ) width formed near grain boundaries under neutron and electron irradiation. The microstructures of Fe-15Cr-xNi (x=15, 20, 25, 30 mass%) alloys that were neutron irradiated at 749 K and electron irradiated at 576 K-824 K were examined. The VDZ widths increased with increasing Ni content in both irradiation experiments, which implies an increase of the vacancy mobility. The vacancy migration energies were estimated from the temperature dependence of the VDZ widths, and the energies were 1.09, 0.97, 0.90, and 0.77 eV for the alloys containing 15, 20, 25, and 30 mass% Ni, respectively. From the obtained energies, the effective vacancy diffusivity and excess vacancy concentration were estimated using the analytical equation of the VDZ width, which quantitatively confirmed the increase of the vacancy mobility with increasing Ni content.
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
Times Cited Count：18 Percentile：86.00(Materials Science, Multidisciplinary)9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 
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 C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 C. This superior strength seemed to be owing to transformation of the matrix from the 
-phase to the 
-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.
Inoue, Toshihiko; Sekio, Yoshihiro; Watanabe, Hideo*
Materia, 58(2), P. 92, 2019/02
For the evaluation of irradiated segregation behavior, Austenite-based stainless steel for the fast reactor, during irradiation was evaluated by utilizing TIARA facility (Irradiate temperature: 600 C, Dose: 100 dpa) was observed by analytical electron microscope (JEM-ARM20FC). As a result of observation, the large-size void is observed in irradiation area, and MX segregation (containing Niobium) is not observed. In un-irradiation area the MX segregation is observed. And it is observed conspicuously that Nickel is segregation on the void surface. By the latest high-performance TEM utilization, these phenomenon are able to visualize. It is expected for the clarification of the irradiation damage and mechanism of void swelling, by the analyzing these phenomenon utilization with the latest high-performance TEM utilization.
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.
Shizukawa, Yuta; Sekio, Yoshihiro; Sato, Isamu*; Maeda, Koji
Proceedings of 54th Annual Meeting of Hot Laboratories and Remote Handling (HOTLAB 2017) (Internet), 5 Pages, 2017/00
Electrochemical corrosion behavior under salt water in a type 304L stainless steel used to a part of BWR core materials was investigated to evaluate the possibility of crevice corrosion occurrence for the fuel assemblies which experienced seawater exposure in Fukushima Daiichi Nuclear Power Plant (1F) accident. Especially, focusing on the upper end plug part having the 304L SS crevice structure, measurement of repassivation potential for crevice corrosion (
) were carried out using the crevice test pieces fabricated by 304L SS plates. From the results, was lower than the spontaneous potential (
) when the conditions of 2500 ppm chloride ion concentration at over 50 C or that of 2500 ppm at over 80 C, which are included in the SFP water quality conditions. Therefore, in the 304L SS parts of the 1F fuel assemblies that experienced seawater exposure, there is a possibility of crevice corrosion occurrence.
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
Times Cited Count：17 Percentile：82.38(Nuclear Science & Technology)Sato, Yutaka*; Kokawa, Hiroyuki*; Fujii, Hiromichi*; Yano, Yasuhide; Sekio, Yoshihiro
Metallurgical and Materials Transactions A, 46(12), p.5789 - 5800, 2015/12
Times Cited Count：17 Percentile：58.52(Materials Science, Multidisciplinary)Dissimilar friction stir welding (FSW) of an 11% Cr ferritic/martensitic stee (PNC-FMS) to 316-grade austenitic stainless steel was attempted with a view to its potential application to the wrapper tubes of next-generation fast reactors. The mechanical properties and microstructure of the resulting welds were systematically examined, which revealed that FSW produces a defect-free stir zone in which material intermixing is notably absent. That is, both steels are separately distributed along a zigzagging interface in the stir zone when PNC-FMS is placed on the retreating side, with the tool plunging at the butt line. This interface did not act as a fracture site during small-sized tensile testing of the stir zone. Furthermore, the microstructure of the stir zone was refined in both the PNC-FMS and 316 stainless steel sides, resulting in improved mechanical properties over the respective base material regions.
Sekio, Yoshihiro; Yoshimochi, Hiroshi; Kosaka, Ichiro; Hirano, Hiroyasu; Koyama, Tomozo; Kawamura, Hiroshi
Proceedings of 52nd Annual Meeting of Hot Laboratories and Remote Handling Working Group (HOTLAB 2015) (Internet), 8 Pages, 2015/09
Due to the Fukushima Daiichi Nuclear Power Plant accident in March 2011, the safe and secure implementations of the decommissioning for Fukushima Daiichi Nuclear Power Plant has been positioned as the urgent tasks in Japan. Japan Atomic Energy Agency has a critical mission of analysing radioactive wastes having generated by the accident for long-term storage and disposal methods. This will be performed in two hot laboratories to be constructed in Okuma Analysis and Research Center at Fukushima Daiichi Nuclear Power Plant site. In one laboratory, radioactive wastes such as rubbles and secondary wastes will be treated, whereas debris such as fuel debris and high dose structural materials will be handled in the other laboratory. The detail considerations for advanced techniques and experimental apparatus to be installed are underway.
Sekio, Yoshihiro; Yamashita, Shinichiro; Sakaguchi, Norihito*; Takahashi, Heishichiro*
Journal of Nuclear Materials, 458, p.355 - 360, 2015/03
Times Cited Count：28 Percentile：90.05(Materials Science, Multidisciplinary)Irradiation-induced void denuded zone (VDZ) formation near grain boundaries was studied to clarify the effects of minor alloying elements on vacancy diffusivity during irradiation in PNC316 steel. The test materials were Fe-15Cr-15Ni steel without additives and PNC316 stainless steel, which contains minor alloying elements. These steels were neutron-irradiated in the experimental fast reactor JOYO and electron-irradiation was also carried out using 1 MeV high voltage electron microscopy. VDZ formation was analyzed from the TEM microstructural observations after irradiation. VDZs were formed near random grain boundaries in both Fe-15Cr-15Ni and PNC316 steels. The VDZ widths in the PNC316 steel were narrower than those for the Fe-15Cr-15Ni steel for all neutron and electron irradiations. The VDZ width analysis implied that the vacancy diffusivity was reduced in PNC316 steel as a result of interaction of vacancies with minor alloying elements.
Sekio, Yoshihiro; Yamagata, Ichiro; Yamashita, Shinichiro; Inoue, Masaki; Maeda, Koji
Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 10 Pages, 2014/10
In the Fukushima Dai-ichi Nuclear Power Plant accident, seawater was temporarily injected into the spent fuel pools since water cooling and feeding functions were lost. For fuel assemblies which experienced seawater immersion, surface corrosion due to seawater constituents and the resultant degradation of mechanical property are of concern. Therefore, in order to assess the integrity of fuel assemblies (especially cladding tubes), the effects of seawater immersion on corrosion behavior and mechanical properties for as-recieved and irradiated Zircaloy-2 cladding tubes were investigated in the present study. As a result, no obvious surface corrosion and no significant degradation in the tensile strength property were observed after both artificial and natural seawater immersion tests for both steels. This suggests that the effects of seawater immersions on corrosion behavior and mechanical property (especially tensile property) for Zircaloy-2 cladding tubes are probably negligible.
Sekio, Yoshihiro; Yamashita, Shinichiro; Sakaguchi, Norihito*; Takahashi, Heishichiro
Materials Transactions, 55(3), p.438 - 442, 2014/03
Times Cited Count：11 Percentile：46.71(Materials Science, Multidisciplinary)In order to perform the comparative evaluations for vacancy diffusivity and flux between a base alloy and modified alloys, the void denuded zones (VDZ) widths were measured from the TEM in-situ observation during electron irradiation in the SUS316L, SUS316L-V and SUS316L-Zr steels. As a result, VDZs with given widths were formed near GBs. Then, the VDZ widths were different depending on steels, and the width was narrower due to addition of minor alloying elements which strongly interact with vacancies. Furthermore, from the analyses of measured VDZ widths in the SUS316L and SUS316L-V steels, the changes of vacancy diffusivity, vacancy flux and excess vacancy concentration were estimated as 0.50, 0.71 and 1.41, respectively. The decreases of vacancy diffusivity and flux during electron irradiation would be due to the interaction of vacancies with added minor elements, while the enhancement of the excess vacancy concentration would be caused by trapping effects due to alloying elements.
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 835C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.
Yano, Yasuhide; Sato, Yutaka*; Sekio, Yoshihiro; Otsuka, Satoshi; Kaito, Takeji; Ogawa, Ryuichiro; Kokawa, Hiroyuki*
Journal of Nuclear Materials, 442(1-3), p.S524 - S528, 2013/09
Times Cited Count：15 Percentile：72.93(Materials Science, Multidisciplinary)Friction stir welding was applied to the wrapper tube materials, 11Cr-ferritic/martensitic steel, intended for fast reactors and defect-free welds were successfully produced. Then, the mechanical and microstructural properties of the friction stir welded steel were investigated. The hardness values of the stir zone were about 550 Hv, and they had hardly any dependence on the rotational speed, although they were much higher than that of the base material. However, tensile strengths and elongations of the stir zones were better at 298 K, compared to those of the base material. These excellent tensile properties were attributable to the fine grain formation during friction stir welding. A part of this study is the result of "Friction stir welding of the wrapper tube materials for Na fast reactors" carried out under the Strategic Promotion Program for Basic Nuclear Research by the Ministry of Education, Culture, Sports, Science and Technology of Japan.
Sato, Yutaka*; Kokawa, Hiroyuki*; Yano, Yasuhide; Sekio, Yoshihiro
Friction Stir Welding and Processing, 7, p.91 - 99, 2013/03
PNC-FMS is a newly developed 11Cr-ferritic/martensitic steel with good swelling resistance, designed for the wrapper tubes of fast reactors. In this study, friction stir welding (FSW) was attempted as a welding process of PNC-FMS because fusion welding of this steel significantly reduced mechanical properties through formation of brittle microstructure. FSW was applied to PNC-FMS at 100 to 300 rpm using Q60 tool, and defect-free welds were obtained. The stir zones had fine microstructure with ferrite and martensite, and the grain size and fraction of martensite increased with rotational speed. Since all welds were overmatched, all transverse tensile specimens were broken at the base material region. The tensile test of the stir zone clarified that the stir zone produced at 100 rpm exhibited higher strengths and elongation than the base material. This study showed that FSW at lower rotational speed produced the stir zone having better mechanical properties in PNC-FMS.
Osaka, Masahiko; Donomae, Takako; Ichikawa, Shoichi; Sasaki, Shinji; Ishimi, Akihiro; Inoue, Toshihiko; Sekio, Yoshihiro; Miwa, Shuhei; Onishi, Takashi; Asaka, Takeo; et al.
Proceedings of 1st Asian Nuclear Fuel Conference (ANFC), 2 Pages, 2012/03
Support system for training and education of future expert in hot laboratories of Oarai-JAEA, named FEETS, is presented. The system has been established based on research results on both characterization of Oarai hot laboratory and user-needs. Various programs under FEETS are also introduced.
Sato, Yutaka*; Kokawa, Hiroyuki*; Yano, Yasuhide; Sekio, Yoshihiro
no journal, ,
no abstracts in English
Sekio, Yoshihiro; Yano, Yasuhide; Sato, Yutaka*; Kokawa, Hiroyuki*
no journal, ,
The applicability of friction-stir welding (FSW) which was already put into practical use with nonferrous-metal materials was examined to the 11Cr-ferrite/martensite steel (PNC-FMS) wrapper material in order to improve the welding performance of fast reactor core materials. In this research, the effect of FSW on mechanical and microstructural properties of the Stir Zone (SZ) was investigated. From the experimental results, there was no remarkable decrease in tensile property of the SZ as compared with the base material even though the hardness of the SZ increased due to the influence of very fine grains. This suggests that a post weld heat treatment which was required on the conventional fusion welding might be unnecessary. This study is the result of "Friction stir welding of the wrapper tube materials for Na fast reactors" carried out under the Strategic Promotion Program for Basic Nuclear Research by the Ministry of Education, Culture, Sports, Science and Technology of Japan.
Yamashita, Shinichiro; Inoue, Masaki; Asaka, Takeo; Sekio, Yoshihiro; Yamagata, Ichiro; Osaka, Masahiko
no journal, ,
R&D program related to integrity assessment of FAs have been done focusing on the influence of seawater injection on corrosion behavior of component materials of these items. Because seawater contains corrosive materials, especially chloride ions, their influence on the long-term integrity of the FAs will be one of the significant concerns during 1F decommissioning work. A basic study plan that supports the estimation of the long-term possible influences of corrosive factors on the component materials of the FAs has been launched. In order to assess the long-term integrity of FAs stored in the common pool of 1F, two kinds of experimental data have to be obtained; one are short-term test data obtained by an immersion test within a real time range and the other are long-term data obtained by an immersion test under conditions that corrosion phenomena are accelerated. In this integrity assessment study, an engineering methodology for the predictive estimation of corrosion of FA component materials is considered and preliminary test results are presented.
