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Sakasegawa, Hideo; Nakajima, Motoki*; Kato, Taichiro*; Nozawa, Takashi*; Ando, Masami*
Materials Today Communications (Internet), 40, p.109659_1 - 109659_8, 2024/08
Times Cited Count：1 Percentile：14.17(Materials Science, Multidisciplinary)Nanometric oxide particles play an important role in improving the creep property of Oxide Dispersion Strengthened (ODS) steels. In our previous research, we examined a microstructural feature known as prior particle boundary (PPB). PPB refers to the surface of mechanically alloyed (MA) powders before consolidation. We revealed that the ODS steel with fine PPBs produced from smaller MA powders, exhibited shorter creep rupture times, compared to that with coarse PPBs produced from larger MA powders. The size of MA powders had an impact on the creep property. In this study, we examined the shape of MA powders, which were non-spherical shapes. Such shapes have the potential to induce anisotropic creep behavior. We conducted small punch creep tests on specimens with two different orientations to study the possible anisotropy. The results revealed that the creep rupture times varied depending on the orientation of specimen, thus indicating anisotropic creep property.
Sakasegawa, Hideo; Nomura, Mitsuo; Sawayama, Kengo; Nakayama, Takuya; Yaita, Yumi*; Yonekawa, Hitoshi*; Kobayashi, Noboru*; Arima, Tatsumi*; Hiyama, Toshiaki*; Murata, Eiichi*
Progress in Nuclear Energy, 153, p.104396_1 - 104396_9, 2022/11
Times Cited Count：1 Percentile：11.70(Nuclear Science & Technology)When dismantling centrifuges in uranium-enrichment facilities, decontamination techniques must be developed to remove uranium-contaminated surfaces of dismantled parts selectively. Dismantled uranium-contaminated parts can be disposed of as nonradioactive wastes or recycled after decontamination appropriate for clearance. previously, we developed a liquid decontamination technique using acidic electrolyzed water to remove uranium-contaminated surfaces. However, further developments are still needed for its actual application. Dismantled parts have various uranium-contaminated surface features due to varied operational conditions, inhomogeneous decontamination using iodine heptafluoride gas, and changes in long-term storage conditions after dismantling. Here, we performed liquid decontamination on specimens with varying uranium-contaminated surfaces cut from a centrifuge made of low-carbon steel. From the results, the liquid decontamination can effectively remove the uranium-contaminated surfaces, and radioactive concentrations fell below the target value within twenty minutes. Although the required time should also depend on dismantled parts' sizes and shapes in their actual application, we demonstrated that it could be an effective decontamination technique for uranium-contaminated steels of dismantled centrifuges.
Sakasegawa, Hideo
ENEKAN, 20, p.20 - 23, 2022/07
no abstracts in English
Kim, B. K.*; Tan, L.*; Sakasegawa, Hideo; Parish, C. M.*; Zhong, W.*; Tanigawa, Hiroyasu*; Kato, Yutai*
Journal of Nuclear Materials, 545, p.152634_1 - 152634_12, 2021/03
Times Cited Count：4 Percentile：36.41(Materials Science, Multidisciplinary)Sakasegawa, Hideo; Tanigawa, Hiroyasu
Fusion Engineering and Design, 109-111(Part B), p.1724 - 1727, 2016/11
Fusion DEMO reactor requires over 11,000 tons of reduced activation ferritic/martensitic steel and it is important to develop the manufacturing technology for producing large-scale components of DEMO blanket with appropriate mechanical properties. In this work, we studied mechanical properties of ferritic/martensitic steel F82H plates with different thicknesses. This is because mechanical properties are generally degraded with increasing production volume and size. As the result, their homogeneity and anisotropy were not significant. However, mass effect was found in their Charpy impact property with increasing plate thickness, i.e. the ductile brittle transition temperature (DBTT) of a 100 mm thick plate was higher than those of the other plates, but its DBTT was still lower than 0
C and comparable to the former heats.
Sakasegawa, Hideo; Tanigawa, Hiroyasu; 2 of others*
Fusion Engineering and Design, 98-99, p.2068 - 2071, 2015/10
DEMO reactor requires over 11,000 tons of reduced activation ferritic/martensitic steel (RAFM). Therefore, it is necessary to develop the manufacturing technology for fabricating such large-scale RAFM with appropriate mechanical properties. In this work, we focused mechanical properties of the F82H-BA12 heat which was melted in a 20 tons electric arc furnace. After the melting followed by forging and hot-rolling, this F82H-BA12 heat was heat-treated in four different conditions to optimize heat treatment conditions, and tensile and Charpy impact tests were then performed. The result of these mechanical tests was compared with that of former F82H heats less than 5 tons, which were melted applying vacuum induction melting, in order to study the effect of using electric furnace.
Hirose, Takanori; Sakasegawa, Hideo; Nakajima, Motoki; Tanigawa, Hiroyasu
Fusion Engineering and Design, 98-99, p.1982 - 1985, 2015/10
As a R&D activity on materials engineering for DEMO blanket in ITER-BA activity, characterization of F82H weld joints prepared with Tungsten-Inert-Gas (TIG) and electron beam (EB) have been investigated. In this work, 50 mm thick plates of F82H were welded using both processes. A similar-metal was employed as a filler for TIG welding. Post-weld-heat-treatment was conducted according to the conditions for Grade 91 defined as ASME P-No.15E, Group No.1. Although the maximum and the minimum hardness of the both joint are similar, the hardness distribution is quite different. The width of EB welds were smaller than that of TIG, and the hardness of EB weld metal was 10% higher than that of TIG. In the TIG welds, the strongest part was heat affected zone (HAZ) heated above phase transformation temperature, Ac1 and the hardness was very similar to the weld metal of EB joint, 280Hv. The hardness of TIG weld metal was around 260 Hv. Both welds demonstrated the smallest hardness, 180 Hv in the HAZ heated below Ac1 temperature. As a investigation of manufacturing process of box fabrication, second EB weld bead was perpendicularly put on the first EB bead. As a result, the second weld did not weaken the HAZ, but reduced the hardness of the weld metal to 260 Hv.
Hirose, Takanori; Nozawa, Takashi; Stoller, R. E.*; Hamaguchi, Dai; Sakasegawa, Hideo; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Enoeda, Mikio; Kato, Yutai*; Snead, L. L.*
Fusion Engineering and Design, 89(7-8), p.1595 - 1599, 2014/10
Times Cited Count：51 Percentile：96.00(Nuclear Science & Technology)The material properties, focusing on the properties used for design analysis were re-assessed and newly investigated for various heats including F82H-IEA. Moreover, irradiation effects on those properties were studied in this work. As for thermal properties, thermal conductivity that has significant impacts on the thermo-hydraulic properties of the blanket was investigated on several heats of F82H including F82H-IEA. According to the measurements, the thermal conductivity falls in the range 28.3
1.1 W/m/K at 293 K. Although this is comparable with that of the other ferritic/martensitic steels, it is 20% lower than the published value for F82H-IEA. The re-assessment on the published value revealed that the thermal diffusivity was over-estimated. As for irradiation effects on the physical properties, electric resistivity was measured after irradiation up to 6 dpa at 573 K and 673 K. The reduction of resistivity in F82H and its welds were 3% and 6%, respectively.
Sakasegawa, Hideo; Tanigawa, Hiroyasu; Ando, Masami
Journal of Nuclear Science and Technology, 51(6), p.737 - 743, 2014/06
Times Cited Count：8 Percentile：48.98(Nuclear Science & Technology)Oxide-dispersion-strengthened (ODS) steels are attractive materials for the fuel cladding of fast reactors and the first-wall material of fusion blanket. High-chromium ferritic ODS steels have better corrosion-resistance properties, but they have poor material workability and anisotropy, making their practical application difficult. In contrast, low-chromium ferritic/martensitic ODS steels have better workability and their anisotropy can be reduced through martensitic transformation. However, their corrosion-resistance properties are poor, compared to high-chromium ferrtic ODS steels. In this work, we developed a corrosion-resistant coating technique for 8Cr ferritic/martensitic ODS steel. The ODS steel was coated with 304 or 430 stainless steel by changing the canning material from mild steel to stainless steel in the conventional material processing procedure and using it as a coating material.
Sakasegawa, Hideo; Tanigawa, Hiroyasu
Journal of Nuclear Materials, 442(1-3), p.S18 - S22, 2013/11
Times Cited Count：1 Percentile：9.76(Materials Science, Multidisciplinary)Through the Broader Approach (BA) activity in Japan, F82H-BA07 heat of 5 tons prepared applying electrosrag remelting (ESR) has been studied as a step toward a larger-scale melting about 20 tons. From the result of elemental mapping images using electron probe microanalysis (EPMA), micro-segregation of at least four metallic elements such as chromium, tungsten, vanadium and manganese was found as stripes parallel to the hot rolling direction. In the case of tungsten segregation, the maximum difference of content was about 1.0 wt% between the observed stripes. This difference could cause differences in nano-metric structures between stripes, and affect mechanical properties. In this presentation, we discuss how much micro-segregation should be decreased considering effects of micro-segregation on nano-metric structures and mechanical properties in addition to the result of optimization of homogenizing condition.
Shiba, Kiyoyuki; Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro
Fusion Engineering and Design, 86(12), p.2895 - 2899, 2011/12
Times Cited Count：45 Percentile：93.85(Nuclear Science & Technology)Aging properties of reduced activation ferritic/martensitic steel F82H was researched at temperature ranging from 400C to 650C up to 100,000 hr. Microstructure, tensile, and Charpy properties were carried out. Laves was found at temperatures between 550 and 650C and M
C carbides were found at the temperatures between 500 and 600C over 10,000 hr. These precipitates caused degradation in toughness, especially at temperatures ranging from 550C to 650C. Tensile properties do not have serious aging effect, except for 650C, which caused large softening even after 10.000 hr. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in DBTT caused by the large Laves phase precipitation at grain boundary was observed in the 650C aging. Laves precipitates at grain boundary also degrades the USE of the aged materials. These aging test results provide F82H can be used up to 30,000 hr at 550C.
Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Sakasegawa, Hideo; Hirose, Takanori; Jitsukawa, Shiro
Fusion Engineering and Design, 86(9-11), p.2549 - 2552, 2011/10
Reduced activation ferritic/martensitic (RAFM) steels are recognized as the primary candidate structural materials for fusion blanket systems, and it is expected to have sound engineering bases, such as fabrication technology and materials database to use RAFM as the structural materials for pressure equipment. It is also important to develop irradiation database and design methodology of fusion neutron irradiated structure to use RAFM as the structural material for fusion neutron irradiated pressure equipment. In the International Fusion Energy Research Centre (IFERC) project in the Broader Approach (BA) activities between EU and Japan, R&D on RAFM steel is underway and these are expected be the bases for DEMO design criteria and licensing. The objective of this paper is to review the BA R&D status of RAFM steel, especially F82H development in Japan. The identified key technical issues for the design and fabrication of DEMO blanket and recent achievements in Japan were introduced.
Sakasegawa, Hideo; Tanigawa, Hiroyasu; Kano, Sho; Enomoto, Masato*
Fusion Engineering and Design, 86(9-11), p.2541 - 2544, 2011/10
Times Cited Count：17 Percentile：74.97(Nuclear Science & Technology)Reduced Activation Ferritic/Martensitic steels (RAFMs) are leading candidates for the structural material of DEMO blanket module. Through the Broader Approach (BA) activity in Japan, the fabrication techniques for the DEMO blanket module has been studied and developed. In the techniques, the development of joining technique is especially important for fabricating the complicated structure of blanket module. In particular, Hot Isostatic Pressing (HIP) has been applied to joining cooling channels with a rectangular cross section. During and after HIP, the structural material are exposed to various heat treatments such as holding at the HIP temperature, following furnace cooling, 2nd normalizing to refine austenite grains, and 2nd tempering. Microstructural evolutions during these various heat treatments should be focused, because they determine the performance of the blanket module. Especially, fine precipitates such as tantalum and vanadium carbides precipitated at high temperatures greatly affect the creep property, the material toughness, and irradiation resistances of RAF as the structural material. In this work, we have studied the stability of fine precipitates in the F82H-BA07 heat (8Cr-2W-V, Ta) during simulated heat treatments of the blanket fabrication.
Sakasegawa, Hideo; Legendre, F.*; Boulanger, L.*; Brocq, M.*; Chaffron, L.*; Cozzika, T.*; Malaplate, J.*; Henry, J.*; de Carlan, Y.*
Journal of Nuclear Materials, 417(1-3), p.229 - 232, 2011/10
Times Cited Count：68 Percentile：97.30(Materials Science, Multidisciplinary)In our past work, the commercial ferrtic Oxide Dispersion Strengthened (ODS) alloy MA957 had at least two types of nanometer-sized oxide particles: non-stoichiometric Y-, Ti-, O-enriched clusters and Y
TiO
particles. The size of the non-stoichiometric clusters was much smaller than that of YTiO particles and it was confirmed that the non-stoichiometric clusters possibly dominate the oxide dispersion strengthening. Therefore, this study dealt with the stability and evolution mechanisms of non-stoichiometric nanoclusters after the annealing (1473K 
1h). This annealing condition was determined considering the actual condition of consolidation processes. After the annealing, most non-stoichiometric Y-, Ti-, O-enriched clusters were stable, but some clusters became YTiO particles with increasing size. The diffusion of yttrium had an important role for the evolution of these oxides.
Ukai, Shigeharu*; Otsuka, Satoshi; Kaito, Takeji; Sakasegawa, Hideo*; Chikada, Nobuyoshi*; Hayashi, Shigenari*; Onuki, Somei*
Materials Science & Engineering A, 510-511, p.115 - 120, 2009/06
Times Cited Count：114 Percentile：96.52(Nanoscience & Nanotechnology)Oxide dispersion strengthened (ODS) ferritic steels, which are the most promising candidate materials for advanced fast reactor fuel elements, have exceptional creep strength at 973 K. The superior creep property of 9Cr-ODS ferritic steels is ascribed to the formation of a nonequilibrium phase, designated as the residual ferrite. The yield strength of the residual ferrite itself has been determined to be as high as 1360 MPa at room temperature from nanoindentation measurements. The creep strength is enhanced by minimizing the number of packet boundaries induced by the martensitic phase transformation. The creep strain occurs by sliding at weaker regions such as at the grain boundaries and packet boundaries. It is found that 9Cr-ODS ferritic steels behave as fiber composite materials comprising the harder residual ferrite and the softer tempered martensite.
Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki; Asayama, Tai; Kim, S.-W.; Ukai, Shigeharu*; Narita, Takeshi*; Sakasegawa, Hideo*
Journal of Nuclear Materials, 386-388, p.479 - 482, 2009/04
Times Cited Count：22 Percentile：78.22(Materials Science, Multidisciplinary)This paper discusses the effects of small portion of Al contamination (
0.1wt%) on the high-temperature strength and microstructure of 9Cr-ODS steel. Increasing Al concentration is shown to provide small reduction of ultimate tensile strength as well as 0.2% proof stress at 973 K and 1073 K accompanied by reduction of elongated grains i.e. residual-
ferrite acting as reinforcement phase. Addition of Al appears to increase the proportion of ferrite phase, which is contrary to general behavior in conventional steels. This unique behavior could be peculiar to the non-equilibrium materials such as mechanically-alloyed alloy. Computer simulation on phase transformation suggests that the fine oxide dispersion in the elongated ferrite could be attributable to the preferential partitioning of Ti and W in ferrite than in austenite at hot-extrusion process at 1423 K.
Sakasegawa, Hideo
JAEA-Research 2007-053, 50 Pages, 2007/11
JAEA-Research-2007-053.pdf:8.31MB
9Cr-ODS (Oxide Dispersion Strengthened) steels developed by JAEA have superior creep properties. The 9Cr ODS steel displaying an excellent creep property is a dual phase steel. The ODS steel is strengthened by the delta ferrite which has a finer dispersion of oxide particles and shows a higher hardness than the martensite. Its creep behavior is very unique and cannot be interpreted by conventional theories of heat resistant steels. Alternative model of creep mechanism was studied using the results of microstructural observations. Based on the alternative creep mechanism model, a novel creep constitutive equation was formulated. In addition to that, modifications of material processing procedures for improving the creep property under irradiation and unirradiation were performed considering procedures for mass production.
Otsuka, Satoshi; Ukai, Shigeharu; Sakasegawa, Hideo; Fujiwara, Masayuki; Kaito, Takeji; Narita, Takeshi
Journal of Nuclear Materials, 367-370(1), p.160 - 165, 2007/08
Times Cited Count：61 Percentile：95.71(Materials Science, Multidisciplinary)This paper describes the effect on creep strength and microstructure of 9Cr-oxide dispersion strengthened martensitic steel (9Cr-ODS steel) brought by the differences in titanium concentration and consolidation temperature. The increase of titanium concentration to 0.30-0.35wt% was shown to provide remarkable improvement of creep strength accompanied by the increase of residual-alpha ferrite. The elevation of hot-extrusion temperature notably degraded the creep strength, however, appeared to increase the volume fraction of residual-alpha ferrite. Creep deformation process of 9Cr-ODS steel was discussed to explain these results based on microstructure observations.
Tanigawa, Hiroyasu; Sakasegawa, Hideo; Ogiwara, Hiroyuki*; Kishimoto, Hirotatsu*; Koyama, Akira*
Journal of Nuclear Materials, 367-370(1), p.132 - 136, 2007/08
Times Cited Count：51 Percentile：93.95(Materials Science, Multidisciplinary)It was previously reported that reduced-activation ferritic/martensitic steels (RAFs), such as F82H-IEA and JLF-1, showed a variety of changes in its mechanical property after neutron irradiation at 573K up to 5dpa, and have possible correlation with precipitation. The effects of irradiation on precipitation were also reported previously. In this study, irradiation effects on precipitation were investigated in detail utilizing ion irradiation in which irradiation condition could be controlled with high accuracy. F82H IEA heat, JLF-1 HFIR heat, and aged F82H-IEA (873K30k h) were used for experiments. The specimens were irradiated at DuET facility, Inst. of Advanced Energy, Kyoto University up to 10 dpa at 573K with 6.4MeV Fe
ion. Cross sectional TEM thin film specimens of ion irradiated region were made utilizing focused ion beam (FIB) processor with micro-sampling system at JAERI. These thin film specimens were made to contain both irradiated region and non-irradiated region beneath irradiated region. Size distribution and aspect ratio of precipitates were analyzed on each region. It turned out that the finer precipitates were dominant in irradiated region of F82H compared to that in non-irradiated region, but fewer and larger precipitates were dominant in irradiated region of JLF-1. These results confirmed the presence of irradiation effects on precipitate evolution even at 573K, which was observed in neutron irradiated RAFs.
Tanigawa, Hiroyasu; Sakasegawa, Hideo; Hashimoto, Naoyuki*; Klueh, R. L.*; Ando, Masami; Sokolov, M. A.*
Journal of Nuclear Materials, 367-370(1), p.42 - 47, 2007/08
Times Cited Count：27 Percentile：84.36(Materials Science, Multidisciplinary)It was previously reported that reduced-activation ferritic/martensitic steels (RAFs), such as F82H-IEA and its heat treatment variant, ORNL9Cr-2WVTa, JLF-1 and 2%Ni-doped F82H, showed a variety of changes in ductile-brittle transition temperature (DBTT) and yield stress after irradiation at 573K up to 5dpa. These differences could not be interpreted solely as an effect of irradiation hardening caused by dislocation loop formation. To address these observations, the precipitation behavior of the irradiated steels was examined by weight analysis, X-ray diffraction analysis and chemical analysis on extraction residues. The results suggested that irradiation affects precipitation as if it was forced to reach the thermal equilibrium state at irradiation temperature 573K, which usually never be achieved by aging. The details of precipitates in the irradiated RAFs were examined to determine their impact on the mechanical properties, which obtained by tensile, Charpy impact, and bend bar toughness tests. Transmission electron microscopy was performed on thin films and extraction replica specimens to analyze the size distribution, chemical composition and crystal structure of precipitates. It turned out that the hardening level normalized by square root of average packet size showed a linear dependence on the increase of extracted precipitate weight. This dependence suggests that the difference in irradiation hardening between RAFs was caused by the different precipitation behavior on packet, block and prior austenitic grain boundaries during irradiation. The simple Hall-Petch law could be applicable to interpret this dependence. Detailed analytical results will be presented and their interpretation discussed.
