Liquid decontamination using acidic electrolyzed water for various uranium-contaminated steel surfaces in dismantled centrifuge

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

https://doi.org/10.1016/j.pnucene.2022.104396

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.

Effects of helium on irradiation response of reduced-activation ferritic-martensitic steels; Using nickel isotopes to simulate fusion neutron response

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

https://doi.org/10.1016/j.jnucmat.2020.152634

Corrosion-resistant coating technique for oxide-dispersion-strengthened ferritic/martensitic steel

Sakasegawa, Hideo; Tanigawa, Hiroyasu; Ando, Masami

Journal of Nuclear Science and Technology, 51(6), p.737 - 743, 2014/06

AA2013-0280.pdf:1.68MB

https://doi.org/10.1080/00223131.2014.894950

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.

Long-term properties of reduced activation ferritic/martensitic steels for fusion reactor blanket system

Shiba, Kiyoyuki; Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro

Fusion Engineering and Design, 86(12), p.2895 - 2899, 2011/12

https://doi.org/10.1016/j.fusengdes.2011.06.005

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 MC 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.

Stability of non-stoichiometric clusters in the MA957 ODS ferrtic alloy

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

https://doi.org/10.1016/j.jnucmat.2010.12.056

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 YTiO 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.

Effects of aluminum on high-temperature strength of 9Cr-ODS steel

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

https://doi.org/10.1016/j.jnucmat.2008.12.147

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.

Irradiation effects on precipitation and its impact on the mechanical properties of reduced-activation ferritic/martensitic steels

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

https://doi.org/10.1016/j.jnucmat.2007.03.167

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.

Nano-mesoscopic structural characterization of 9Cr-ODS martensitic steel for improving creep strength

Otsuka, Satoshi; Ukai, Shigeharu; Sakasegawa, Hideo; Fujiwara, Masayuki; Kaito, Takeji; Narita, Takeshi

Journal of Nuclear Materials, 367-370(1), p.160 - 165, 2007/08

https://doi.org/10.1016/j.jnucmat.2007.03.004

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.

Microstructure property analysis of HFIR-irradiated reduced-activation ferritic/martensitic steels

Tanigawa, Hiroyasu; Hashimoto, Naoyuki*; Sakasegawa, Hideo*; Klueh, R. L.*; Sokolov, M. A.*; Shiba, Kiyoyuki; Jitsukawa, Shiro; Koyama, Akira*

Journal of Nuclear Materials, 329-333(1), p.283 - 288, 2004/08

https://doi.org/10.1016/j.jnucmat.2004.04.053

Reduced-activation ferritic/martensitic steels (RAFs) were developed as candidate structural materials for fusion power plants. In a previous study, it was reported that ORNL9Cr-2WVTa and JLF-1 (Fe-9Cr-2W-V-Ta-N) steels showed smaller ductile-brittle transition temperature (DBTT) shifts compared to IEA modified F82H (Fe-8Cr-2W-V-Ta) after neutron irradiation up to 5 dpa at 573K. This difference in DBTT shift could not be interpreted as an effect of irradiation hardening, and it is also hard to be convinced that this difference was simply due to a Cr concentration difference. To clarify the mechanisms of the difference in Charpy impact property between these steels, various microstructure analyses were performed.

Analysis of extraction residue of HFIR 11J-irradiated RAFs

Tanigawa, Hiroyasu; Sakasegawa, Hideo*; Zinkle, S. J.*; Klueh, R. L.*; Koyama, Akira*

Fusion Materials Semiannual Progress Report for the Period Ending (DOE/ER-0313/35), p.30 - 32, 2004/04

Extraction residue was made from several HFIR 11J-irradiated RAFs, and the mass change was measured to investigate the irradiation-enhanced change in precipitation. Two different types of filter with coarse and fine pores were used in order to separate the difference of irradiation effects between larger and smaller precipitates. Unirradiated specimens were examined as well. Results suggest that during irradiation the mass of larger precipitates increased in F82H-IEA, Ni-doped F82H, JLF-1 and ORNL9Cr, fine precipitates disappeared in JLF-1, and fine precipitates increased in Ni-doped F82H.