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Journal Articles

Effect of nickel concentration on radiation-induced diffusion of point defects in high-nickel Fe-Cr-Ni model alloys during neutron and electron irradiation

Sekio, Yoshihiro; Sakaguchi, Norihito*

Materials Transactions, 60(5), p.678 - 687, 2019/05

 Times Cited Count:2 Percentile:25.87(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.

Journal Articles

Ultra-high temperature creep rupture and transient burst strength of ODS steel claddings

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:7 Percentile:85.54(Materials Science, Multidisciplinary)

9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 $$^{circ}$$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 $$^{circ}$$C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 $$^{circ}$$C. This superior strength seemed to be owing to transformation of the matrix from the $$alpha$$-phase to the $$gamma$$-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.

Journal Articles

Austenite-based stainless steel irradiation behavior of the precipitate and void swelling

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 $$^{circ}$$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.

Journal Articles

Evaluation of irradiation-induced point defect migration energy during neutron irradiation in modified 316 stainless steel

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.

Journal Articles

Electrochemical corrosion tests for core materials utilized in BWR under conditions containing seawater

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 ($$E_{rm R,CREV}$$) were carried out using the crevice test pieces fabricated by 304L SS plates. From the results, $$E_{rm R,CREV}$$ was lower than the spontaneous potential ($$E_{rm SP}$$) when the conditions of 2500 ppm chloride ion concentration at over 50 $$^{circ}$$C or that of 2500 ppm at over 80 $$^{circ}$$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.

Journal Articles

Tensile properties and hardness of two types of 11Cr-ferritic/martensitic steel after aging up to 45,000 h

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

BB2015-1728.pdf:1.04MB

 Times Cited Count:9 Percentile:75.69(Nuclear Science & Technology)

Journal Articles

Mechanical properties and microstructure of dissimilar friction stir welds of 11Cr-ferritic/martensitic steel to 316 stainless steel

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:12 Percentile:60.62(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.

Journal Articles

Program of the analysis and research laboratory for Fukushima-Daiichi and advanced techniques to be applied in the laboratory

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.

Journal Articles

Void denuded zone formation for Fe-15Cr-15Ni steel and PNC316 stainless steel under neutron and electron irradiations

Sekio, Yoshihiro; Yamashita, Shinichiro; Sakaguchi, Norihito*; Takahashi, Heishichiro*

Journal of Nuclear Materials, 458, p.355 - 360, 2015/03

 Times Cited Count:13 Percentile:81.43(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.

Journal Articles

Seawater immersion tests of irradiated Zircaloy-2 cladding tube

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.

Journal Articles

Effect of additional minor elements on accumulation behavior of point defects under electron irradiation in austenitic stainless steels

Sekio, Yoshihiro; Yamashita, Shinichiro; Sakaguchi, Norihito*; Takahashi, Heishichiro

Materials Transactions, 55(3), p.438 - 442, 2014/03

 Times Cited Count:8 Percentile:46.14(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.

JAEA Reports

Evaluation of irradiation behavior on oxide dispersion strengthened (ODS) steel claddings irradiated in Joyo/CMIR-6

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 835$$^{circ}$$C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.

Journal Articles

Mechanical properties of friction stir welded 11Cr-ferritic/martensitic steel

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:12 Percentile:73.42(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.

Journal Articles

Effect of welding parameters on microstructure and mechanical properties of a friction stir welded 11Cr-ferritic/martensitic steel

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.

Journal Articles

Support system for training and education of future expert at PIE Hot Laboratories in Oarai JAEA; FEETS

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.

Oral presentation

Swelling properties of neutron-irradiated ceramics in JOYO

Sekio, Yoshihiro; Tachi, Yoshiaki; Watanabe, Masashi*; Shikama, Tatsuo*

no journal, , 

no abstracts in English

Oral presentation

Effect of nature of grain boundary on void distribution during irradiation in austenitic stainless steels

Sekio, Yoshihiro; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Nishinoiri, Kenji; Takahashi, Heishichiro*

no journal, , 

In this study, microstructural observation of Fe-15Cr-15Ni ternary model alloy and PNC316 irradiated in JOYO was performed to clarify the defect accumulation process in the vicinity of grain boundary and also the relationship between grain boundary coherency and the void structure developing process was evaluated. As the result of microstructural observation, the followings are obtained; void denuded zone (VDZ) was developed in the vicinity of the low symmetry grain boundary of Fe-15Cr-15Ni whereas no VDZ developed in the vicinity of high symmetry grain boundary. In the case of PNC316, narrow VDZ was developed only at the low symmetry grain boundary.

Oral presentation

Effects of nature of grain boundary on void distribution during irradiation in austenitic stainless steels

Sekio, Yoshihiro; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Nishinoiri, Kenji; Takahashi, Heishichiro

no journal, , 

The process of the irradiation-induced defects accumulation during irradiation has not been clarified yet, namely what microstructural factors affect defect accumulation process. For this reason, we investigated the void distribution near grain boundaries by focusing on grain boundaries as a microstructural factor in Fe-15Cr-15Ni model alloy and PNC316 stainless steel used for the nuclear fast reactor. It was clarified that the void denuded zone were formed near only a random grain boundary under neutron irradiation in Fe-15Cr-15N alloy and PNC316 stainless steel as being observed in the electron-irradiated austenitic stainless steels. We discussed the difference of the void denuded zone distribution in both steels under low and high fluencie. These results suggest that the grain boundary characteristics and its effect on void denuded zone formation will play a role for the void swelling under neutron irradiation in austenitic stainless steels.

Oral presentation

Future expert educational/training support system at Fuels and Materials Department of JAEA, 1; Basic concept and preparation of introductive apparatuses

Asaka, Takeo; Osaka, Masahiko; Donomae, Takako; Ichikawa, Shoichi; Sasaki, Shinji; Ishimi, Akihiro; Inoue, Toshihiko; Sekio, Yoshihiro; Miwa, Shuhei; Onishi, Takashi; et al.

no journal, , 

Activities on a system development for education and training of future experts in the nuclear field thorough experiments at PIE facilities are given. Basic concept of the system and acceptance-support apparatus are introduced.

Oral presentation

Microstructures and mechanical properties of friction stir welded 11Cr-ferritic/martenstic steel

Sato, Yutaka*; Kokawa, Hiroyuki*; Yano, Yasuhide; Sekio, Yoshihiro

no journal, , 

no abstracts in English

52 (Records 1-20 displayed on this page)