Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 24

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Corrosion behavior of ODS steels with several chromium contents in hot nitric acid solutions

Tanno, Takashi; Takeuchi, Masayuki; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Materials, 494, p.219 - 226, 2017/10

BB2016-1307.pdf:0.6MB

 Times Cited Count:12 Percentile:86.92(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) steel cladding tubes have been developed for fast reactors. 9 chromium ODS and 11Cr-ODS tempered martensitic steels are prioritized for the candidate material in research being carried out at JAEA. In this work, fundamental immersion tests and electro-chemical tests of 9 to 12Cr-ODS steels were systematically conducted in various nitric acid solutions at 95$$^{circ}$$C. The corrosion rate exponentially decreased with effective solute chromium concentration (Cr$$_{eff}$$) and nitric acid concentration. Addition of oxidizing ions also suppressed the corrosion rate. According to polarization curves and surface observations in this work, the combination of low Cr$$_{eff}$$ and dilute nitric acid could not prevent the active dissolution at the beginning of immersion, and the corrosion rate was high. In comparison, higher Cr$$_{eff}$$, concentrated nitric acid and addition of oxidizing ions helped to prevent the active dissolution, and suppressed the corrosion rate.

Journal Articles

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.

Journal of Nuclear Materials, 487, p.229 - 237, 2017/04

 Times Cited Count:27 Percentile:97.79(Materials Science, Multidisciplinary)

Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400$$^{circ}$$C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200$$^{circ}$$C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200$$^{circ}$$C. This degradation was attributed to grain boundary sliding deformation with $$gamma$$/$$delta$$ transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 $$^{circ}$$C unlike the other tested materials. Present study includes the result of "R&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).

Oral presentation

Characterization of melt-solidified (U, Gd, Zr)O$$_{2-x}$$ as simulated corium debris

Morimoto, Kyoichi; Hiroka, Shun; Akashi, Masatoshi; Watanabe, Masashi

no journal, , 

The influence of Gd on characteristics of debris is important for removing the debris from the reactors of Fukushima Daiichi Nuclear Power Plant because subassemblies of nuclear fuels containing Gd$$_{2}$$O$$_{3}$$ were loaded in the some reactor cores. Additionally, it is important to assess the distribution state of Gd from the anxiety of re-criticality caused by the relocation of debris while removing them. In this study, sintered pellets of (U$$_{0.95-y}$$Gd$$_{0.05}$$Zr$$_{y}$$)O$$_{2-x}$$ (y=0,0.5, 2-x=1.989-2.000) were melted and solidified to prepare specimens of simulated corium debris. Phase states and fundamental properties of them were evaluated.

Oral presentation

The Effect of Am on oxygen potential of MOX fuel at high temperatures

Matsumoto, Taku; Morimoto, Kyoichi; Kato, Masato; Sunaoshi, Takeo*

no journal, , 

Oral presentation

Phase state estimation of metallic inclusions in simulated corium debris

Akashi, Masatoshi; Morimoto, Kyoichi

no journal, , 

Oral presentation

Oral presentation

Oxygen self and chemical diffusion coefficients in (U, Pu)O$$_{2}$$

Watanabe, Masashi; Kato, Masato; Sunaoshi, Takeo*

no journal, , 

Diffusion phenomena of oxygen in mixed oxide fuel are especially important in understanding fuel behaviour. The reason is that sintering, evaporation, oxygen redistribution and behaviour of fission products are essentially involved in the oxygen diffusion process. Thus, the purpose of this work is to precisely measure the oxygen self and chemical diffusion coefficients in (U, Pu)O$$_{2}$$ at high temperatures and to evaluate the relationship between both coefficients.

Oral presentation

Japan's efforts to develop performance assessment models for waste glass corrosion

Mitsui, Seiichiro

no journal, , 

In order to develop robust performance assessment models on waste glass corrosion, JAEA has been conducting experimental studies and has been preparing an information basis regarding the near-field processes under disposal conditions. To determine silicon migration parameters such as distribution coefficient of Si in buffer material, we conducted percolation type migration experiments and batch sorption experiments using Si-32 as radioactive tracer for Kunigel-V1 as a joint project with SCKCEN. The values of the apparent diffusion coefficient and the distribution coefficient of dissolved silica in Kunigel-V1 are estimated to be 3.4 $$times$$ 10$$^{-13}$$ $$m^{2}$$/s and 0.5 $$m^{3}$$/kg on average, respectively. The preparation of the information basis has been conducted as a part of joint project with NUMO. Based on the information basis, we developed a preliminary model of the glass corrosion considering the near-field processes and conducted sensitivity analyses for selected processes.

Oral presentation

Oral presentation

High temperature creep properties of ODS steel cladding for evaluating severe accident

Kato, Shoichi; Furukawa, Tomohiro; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Oka, Hiroshi; Inoue, Toshihiko; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.

no journal, , 

Oxide dispersion strengthened (ODS) steel is a prime candidate for cladding tubes of Japan Sodium-cooled Fast Reactor (JSFR) due to the high temperature and radiation resistances. One of the safety design of JSFR for Design Extension Condition (DEC) is the control of severe plant conditions, including prevention of severe accidents and mitigation of severe-accident consequences. Therefore, it is necessary to acquire the mechanical properties at ultra-high temperature conditions for core materials to evaluate safety design. There are, however, no data for ODS claddings at ultra-high temperature condition for the reflecting to the design criteria. In this study, creep rupture tests of 9Cr-ODS, 12Cr-ODS and FeCrAl-ODS steel claddings have been done at elevated temperatures, and the effect of minor elements such as Al, Zr and O on the mechanical strength and the creep rupture curve for the safety design were evaluated. The effect of minor elements was estimated based on the data at 700$$^{circ}$$C and 1000$$^{circ}$$C. As the results, it was confirmed that the addition of Zr had an effect on the improvement of creep strength at elevated temperature for the FeCrAl-ODS steel claddings.

Oral presentation

Drying characteristics of simulated debris in a pretreatment process for dry storage

Nakayoshi, Akira; Suzuki, Seiya; Okamura, Nobuo; Watanabe, Masayuki; Koizumi, Kenji

no journal, , 

Treatment policies for debris from Fukushima Daiichi Nuclear Power Plant is not decided, however, any policies may include medium and long term storages of debris. Dry storages may be desirable in terms of costs and handlings, but it is necessary to assess generating hydrogen during storages due to radiolysis of accompanied water with debris before debris storages. Al$$_{2}$$O$$_{3}$$, SiO$$_{2}$$, ZrO$$_{2}$$, UO$$_{2}$$ and cement paste pellets as simulated debris were prepared, which have some porosities and pore. Weight changes of wet samples were measured at various drying temperatures (200, 300, and 1000 $$^{circ}$$C) using a thermal gravity measurement, under helium gas flow (50 cc/min) or reduced pressure conditions (reducing pressure rate: 200 Pa in 30 min.). From the results, drying curves were evaluated.

Oral presentation

Material modification and nanostructuring by swift heavy ions

Ishikawa, Norito

no journal, , 

Radiation damage due to irradiation with swift heavy ions (SHI) with the energy above 1 MeV/u has many different features compared to that due to irradiation with electrons and neutrons. One of the well-known phenomena related to SHI irradiation is formation of ion tracks. An ion track is a cylindrical region where the material near ion-path is locally modified in nanometric scale. Ion tracks are of great interest in a wide variety of research fields including nuclear materials science, physics of ion-solid interaction, nanotechnology, archaeology and so on. Mechanism of ion track formation in inorganic materials has always been one of the central and intriguing subjects in the SHI research community. It is still challenging to untangle the related problems. In this plenary talk, I would like to concentrate on the vital part of the recent advancement so that the audience can understand the logical pathway of the latest research works.

Oral presentation

Experimental studies on Cs chemisorption on reactor structural materials and their behaviour at high temperature

Suzuki, Eriko; Di Lemma, F. G.; Nakajima, Kunihisa; Yamashita, Shinichiro; Osaka, Masahiko

no journal, , 

In order to clarify the re-vaporization behavior of cesium (Cs) chemisorbed compounds which formed onto reactor structural materials during Severe Accident (SA), Cs chemisorbed samples were reheated at 1000$$^{circ}$$C and then microstructural analysis of the chemisorbed samples was conducted. In the case of stainless steel containing Mo, Cs-Mo-O compounds were formed on surface, together with major Cs-Fe-Si-O compounds, and re-vaporized easier than Cs-Fe-Si-O compounds at 1000$$^{circ}$$C.

Oral presentation

A First-principles study on thermal conductivity of actinide dioxides

Nakamura, Hiroki; Machida, Masahiko

no journal, , 

Actinide dioxides, such as UO$$_{2}$$ and PuO$$_{2}$$, are the main components of nuclear fuel. However, the determination of their properties through experiments is not easy owing to limitations associated with their handling. In such cases, numerical simulations are effective for the evaluation of the properties of actinide dioxides. So far, we have evaluated thermal properties of actinide dioxides based on the ground-state calculation using first-principles density functional theory (DFT) and successfully reproduced the observed quantities such as heat capacity. In this paper, we apply this calculation method to thermal-conductivity estimation. The calculated thermal conductivities agree well with measurements between room temperature and 1800 K. In conclusion, our calculation method is available to evaluate thermal conductivity of actinide dioxides and can contribute the development of nuclear fuels.

Oral presentation

Mechanical properties of Urania-Zirconia solid solutions with tetragonal and monoclinic structures

Ikeuchi, Hirotomo; Yano, Kimihiko; Ogino, Hideki; Matsunaga, Junji*

no journal, , 

Mechanical properties (micro-hardness, elastic modulus, and fracture toughness) of fuel debris are essential information for defueling work in the Fukushima Daiichi NPP. The Urania-Zirconia solid solution, (U,Zr)O$$_{2}$$, is expected to be separated into cubic phase with low Zr contents and tetragonal or monoclinic phase with high Zr contents during cooling process. In this study, the properties of tetragonal and monoclinic phase are investigated. The (U$$_{1-x}$$Zr$$_{x}$$)O$$_{2.0}$$ samples (x=0.85 for tetragonal phase and 0.95 for monoclinic phase) were prepared by sintering the compacted mixture of UO$$_{2}$$ and ZrO$$_{2}$$ powders at maximum 2673 K. The micro-hardness of samples was lower than what has been expected from trend of cubic phase. The elastic modulus was comparable with cubic phase. The fracture toughness of the tetragonal phase was higher than the other two phases. The stress-induced martensitic transformation around the indent is expected to increase the fracture toughness.

Oral presentation

Characterization of core melt concrete interface region examined by light concentrating heating technique

Sudo, Ayako; Takano, Masahide; Onozawa, Atsushi

no journal, , 

To characterize the reaction layers around core melt/concrete interface, MCCI experiments by using a light-concentrating technique was performed. As the main constituents of the core-melt, powder mixtures of ZrO$$_{2}$$, Zr, (U,Zr)O$$_{2}$$, stainless steel (SS), and B$$_{4}$$C with various compositions were compacted into tablets. The tablet was placed on a concrete. Light from a lamp was concentrated on the tablet, and the vertical cross-section of the solidified sample was determined by XRD and SEM/EDX. The analyses identified 4 layers from top to bottom; (a) as-melted (U,Zr)O$$_{2}$$ particles and silicate glass with U, (b) the silicate glass with U, (c) imperfectly melted concrete, and (d) dehydrated concrete. Unoxidized metal particles (Fe-Ni-Cr) also precipitated. Gd$$_{2}$$O$$_{3}$$, Mo-Ru-Rh-Pd alloy, and sea salt were also added in the tablet. In this case Gd was included in both (U,Zr)O$$_{2}$$ and silicate glass, Mo and the platinum group elements formed alloys with Fe-Ni-Cr, and S originating from sea salt resulted in precipitation of FeS-type sulfide in the alloy.

Oral presentation

Effects of boron on revaporization of iodine and cesium compounds in a severe accident condition

Miwa, Shuhei; Shinada, Masanori; Osaka, Masahiko; Sugiyama, Tomoyuki; Maruyama, Yu

no journal, , 

In order to acquire the data on fission product chemical behavior during transport in a reactor for the improvement of source term evaluation method, we performed the chemical reaction tests of cesium (Cs) and iodine (I) deposits and boron oxide (B$$_{2}$$O$$_{3}$$) vapor/aerosol using the apparatus which can simulate temperature conditions of reactor coolant system under a sever accident. The volatile I compounds were formed by the reaction of B$$_{2}$$O$$_{3}$$ vapor/aerosol and deposit, and significant amount of I was revaporized from the deposit.

Oral presentation

Progress of R&D and remaining subjects on materials degradation in severe accidents

Nagase, Fumihisa

no journal, , 

JAEA conducts R&D to support the decommissioning at the Fukushima Daiichi NPS and to contribute improvement of the LWR safety in the frame of domestic and international collaborations as well as the own projects. The R&D mostly focuses on the phenomena in BWRs and covers various issues related to materials degradation in severe accidents. In parallel, JAEA has the research activity to establish technical basis for practical use of accident tolerant fuel (ATF) components in existing LWRs. The preliminary computer code analyses showed necessary material data and subjects to design the ATF components.

Oral presentation

Phases and morphology in the simulated MCCI products prepared by arc melting method

Takano, Masahide; Onozawa, Atsushi; Sudo, Ayako

no journal, , 

To understand the characteristics of MCCI products in Fukushima Daiichi Nuclear Power Station, the simulated MCCI products in laboratory scale were prepared by arc melting of compacted powder mixtures of core materials and concrete. Stainless steel, boron carbide, metallic zirconium, (U,Zr)O$$_{2}$$, GdO$$_{1.5}$$, and platinum group elements were selected as the core materials. Phases, morphology, and micro hardness were analyzed on cross-section of the solidified specimens. The specimens consisted of oxide part (MO$$_{2}$$ corium and silicate glass or Al-Ca-O) and metallic part (alloys and borides). The phase relationships in the MCCI products were found to be dominated by the initial concrete/Zr mixing ratio, because the dehydration of concrete is the main oxidation factor and the metallic zirconium acts as a strong reductant. Micro hardness of main phases are 7 GPa for silicate glass, 13-15 GPa for (U,Zr,Gd,Ca)O$$_{2}$$ corium, and 25 GPa for ZrB$$_{2}$$ and ferrous borides, respectively.

Oral presentation

Current status of design, R&D toward ADS target irradiation facility, TEF-T

Okubo, Nariaki; Saito, Shigeru; Obayashi, Hironari; Sasa, Toshinobu

no journal, , 

In the ADS plant, the target window material will be heavily irradiated under the severe condition. Degradation of the mechanical properties, size change of the components and erosion/corrosion of the material surface after irradiation in the LBE flow should be suppressed within a range permissible for the ADS system design. Material irradiation experiments in the transmutation experimental facility (TEF-T), which is planned in JAEA, will realize the first ADS plant in near future. Overview of the material irradiation and PIE plan by using TEF-T including R&D of LBE corrosion test loop and elemental key technology toward TEF-T and ADS will be shown in this presentation. In order to select candidate materials prior to TEF-T irradiation, triple ion irradiations of Fe, He and H ions were conducted for ferritic/martensitic steels, T91 in TIARA. The swelling behavior of T91 irradiated at temperature from 350 to 550$$^{circ}$$C will be also discussed.

24 (Records 1-20 displayed on this page)