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Journal Articles

Effect of nitrogen concentration on nano-structure and high-temperature strength of 9Cr-ODS steel

Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji

Nuclear Materials and Energy (Internet), 16, p.230 - 237, 2018/08

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

First-principles calculations of interaction between solutes and dislocations in tungsten

Tsuru, Tomohito; Suzudo, Tomoaki

Nuclear Materials and Energy (Internet), 16, p.221 - 225, 2018/08

Changes in mechanical properties due to transmutation products (Re and Os) in W alloys is a central issue for plasma-facing materials in fusion reactors. We implemented density function theory calculations to investigate the effect of Re, Os, and other 5d solutes on dislocation core structure and motion associated with plastic deformation. Ir, Pt, Au, and Hg solutes show strong attractive interactions with screw dislocations, causing solution strengthening by the pinning mechanism. On the other hand, Hf, Ta, and Re cause softening by facilitating dislocation motion around solutes. This prediction corresponds well with the experimental observation of softening behavior in W-Re alloys.

Journal Articles

Ion irradiation effects on FeCrAl-ODS ferritic steel

Kondo, Keietsu; Aoki, So; Yamashita, Shinichiro; Ukai, Shigeharu*; Sakamoto, Kan*; Hirai, Mutsumi*; Kimura, Akihiko*

Nuclear Materials and Energy (Internet), 15, p.13 - 16, 2018/05

 Times Cited Count:5 Percentile:29.78(Nuclear Science & Technology)

Radiation hardening and microstructural evolution of ion irradiated 12Cr-6Al ODS ferritic steel was studied. Ion irradiation experiments were performed using Fe ions up to the nominal displacement damage of 20 dpa at the irradiation temperature was 300$$^{circ}$$C. The monotonical increase of radiation hardening with dose was confirmed by experimentally obtained hardness data. The radiation hardening was also calculated theoretically by introducing the microstructural character examined by TEM into the dispersed barrier hardening model. The results showed a good agreement with the experimentally obtained data up to 5 dpa, while a slight discrepancy was found between theoretical and experimental hardness values at 20 dpa. Radiation hardening was mainly caused by irradiation-induced defect clusters below the irradiation dose of 5 dpa. As the irradiation dose increased toward 20 dpa, an additional influence of the radiation appeared, which was assumed to be induced by $$alpha$$' phase transformation.

Journal Articles

Implementation of a low-activation Au-In-Cd decoupler into the J-PARC 1 MW short pulsed spallation neutron source

Teshigawara, Makoto; Ikeda, Yujiro; Oi, Motoki; Harada, Masahide; Takada, Hiroshi; Kakishiro, Masanori*; Noguchi, Gaku*; Shimada, Tsubasa*; Seita, Kyoichi*; Murashima, Daisuke*; et al.

Nuclear Materials and Energy (Internet), 14, p.14 - 21, 2018/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

We developed an Au-In-Cd (AuIC) decoupler material to reduce induced radioactivity instead of Ag-In-Cd one, which has a cut off energy of 1eV. In order to implement it into an actual moderator-reflector assembly, a number of critical engineering issues need to be resolved with regard to large-sized bonding between AuIC and A5083 alloys by the hot isostatic pressing process. We investigated this process in terms of the surface conditions, sizes, and heat capacities of large AuIC alloys. We also show a successful implementation of an AuIC decoupler into a reflector assembly, resulting in a remarkable reduction of radioactivity by AuIC compared to AIC without sacrificing neutronic performance.

Journal Articles

Radioactive contamination of several materials following the Fukushima Daiichi Nuclear Power Station accident

Koma, Yoshikazu; Shibata, Atsuhiro; Ashida, Takashi

Nuclear Materials and Energy (Internet), 10, p.35 - 41, 2017/01

 Times Cited Count:10 Percentile:10.91(Nuclear Science & Technology)

Due to the accident of Fukushima Daiichi Nuclear Power Station (NPS) of Tokyo Power Electric Co. in 2011, radioactive nuclides were released to environment and heavily contaminated the NPS site itself. Analytical concentration of radionuclides for contaminated materials including the accumulated water, rubble, soil and vegetation which already opened were utilized to discuss contamination behavior of some radionuclides. The concentration ratio of a radionuclide to that of $$^{137}$$Cs was normalized with composition of the damaged reactor fuels referred to as "transport ratio" to investigate elemental behavior. The transport ratio of Sr to rubble and soil compared with which was assumed to be via air was approximately 10$$^{-2}$$ to 10$$^{-3}$$, and the successive transport to the accumulated water was comparable to Cs except for the initial release. Transport of Pu, Am and Cm were scarce relative to Cs regardless of transport process by means of air and water, whereas those of iodine, selenium and tellurium were comparable or greater. Contamination with $$^{3H}$$ and $$^{14}$$C was independent of $$^{137}$$Cs, $$^{90}$$Sr and TRU nuclides and would be a result of different transport process.

Journal Articles

Effect of thermo-mechanical treatments on nano-structure of 9Cr-ODS steel

Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Onuma, Masato*

Nuclear Materials and Energy (Internet), 9, p.346 - 352, 2016/12

 Times Cited Count:10 Percentile:18.31(Nuclear Science & Technology)

Journal Articles

Dissolution behavior of lithium compounds in ethanol

Furukawa, Tomohiro; Hirakawa, Yasushi; Kondo, Hiroo; Kanemura, Takuji

Nuclear Materials and Energy (Internet), 9, p.286 - 291, 2016/12


 Times Cited Count:2 Percentile:69.72(Nuclear Science & Technology)

In order to exchange the components which received irradiation damage during the operation at the International Fusion Materials Irradiation Facility, the adhered lithium, which is partially converted to lithium compounds such as lithium oxide and lithium hydroxide, should be removed from the components. In this study, the dissolution experiments of lithium compounds (lithium nitride, lithium hydroxide, and lithium oxide) were performed in a candidate solvent, allowing the clarification of time and temperature dependence. Based on the results, a cleaning procedure for adhered lithium on the inner surface of the components was proposed.

Journal Articles

Effect of dissolved gas on mechanical property of sheath material of mineral insulated cables under high temperature and pressure water

Takeuchi, Tomoaki; Nakano, Hiroko; Uehara, Toshiaki; Tsuchiya, Kunihiko

Nuclear Materials and Energy (Internet), 9, p.451 - 454, 2016/12

 Times Cited Count:1 Percentile:83.2(Nuclear Science & Technology)

no abstracts in English

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


 Times Cited Count:7 Percentile:31.55(Nuclear Science & Technology)

Journal Articles

Strength anisotropy of rolled 11Cr-ODS steel

Tanno, Takashi; Yano, Yasuhide; Oka, Hiroshi; Otsuka, Satoshi; Uwaba, Tomoyuki; Kaito, Takeji

Nuclear Materials and Energy (Internet), 9, p.353 - 359, 2016/12


 Times Cited Count:4 Percentile:47.23(Nuclear Science & Technology)

Materials for core components of fusion reactors and fast reactors, such as blankets and fuel cladding tubes, must be excellent in high temperature strength and irradiation resistance because they will be exposed to high heat flux and heavy neutron irradiation. Oxide dispersion strengthened (ODS) steels have been developing as the candidate material. Japan Atomic Energy Agency (JAEA) have been developing 9 and 11 Chromium (Cr) ODS steels for advanced fast reactor cladding tubes. The JAEA 11Cr-ODS steels were rolled in order to evaluate their anisotropy. Tensile tests and creep tests of them were carried out at 700 $$^{circ}$$C in longitudinal and transverse orientation. The anisotropy of tensile strength was negligible, though that of creep strength was distinct. The observation results and chemical composition analysis suggested that the cause of the anisotropy in creep strength was prior powder boundary including Ti-rich precipitates.

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