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Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji; Ito, Chikara
no journal, ,
JAEA has been developing ODS steels for the high burnup fuel cladding tubes of sodium-cooled fast reactors (SFR). Evaluation of sodium environmental effects is important since the outer surface of SFR fuel cladding tubes are exposed to high temperature flowing sodium and the tube wall is very thin. In this study, the numerical calculations were conducted based on thermodynamics and rate theory for understanding and predicting Cr dissolution behaviors of Fe-Cr steel in flowing sodium. The calculation results indicated that Cr concentration of steel surface gradually deceased with time, and approached to a unique value no matter what Cr concentration the steel contains in initial stage. Increasing flow velocity shortened the time for surface Cr concentration approaching the converged value. In the presentation, the calculated results will be compared to experimentally measured data, and discussions will be conducted to improve the Cr dissolution model constructed in this study.
Kaito, Takeji; Yano, Yasuhide; Shizukawa, Yuta; Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi
no journal, ,
Itagaki, Wataru; Saito, Hiroto; Yamamoto, Masaya; Takamatsu, Misao; Maeda, Shigetaka
no journal, ,
JAEA applied for inspections of Joyo under the new regulation on March 30, 2017. It is undergoing NRA's safety review against core modification to MK-IV 100 MW irradiation core, measurement to natural disaster and BDBAs. The MK-IV core will have the maximum fast neutron flux (E 
0.1 MeV) of 10
n/cm
s. This is a high-level fast neutron flux of the irradiation test facilities in the world. Joyo can further enhance its capabilities by tailoring neutron spectrum, decreasing or increasing the irradiation temperature. For an experiment, it is important to evaluate neutron flux, dpa and temperature. Joyo provides accurate evaluation of irradiation conditions based on calculations verified by previous characterization tests and measurement using dosimetry technique. In this presentation, measure to new regulation, schedule for restart and irradiation capability of Joyo are introduced.
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Sekio, Yoshihiro; Otsuka, Satoshi; Kaito, Takeji
no journal, ,
no abstracts in English
Shizukawa, Yuta; Sekio, Yoshihiro; Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Tachi, Yoshiaki; Otsuka, Satoshi; Kaito, Takeji
no journal, ,
In JAEA's Materials Monitoring Facility (MMF), various post-irradiation examination (PIEs) have been carried out on the long-life core, structural and absorber materials irradiated in the experimental fast reactor Joyo. The PIEs have revealed many important technological insights, i.e. the effects of irradiation on mechanical properties, microstructures, and physical properties of fast reactor materials such as modified austenitic steel, radiation-resistant ferritic steels including oxide dispersion strengthened (ODS) steel, B
C, and so on. The test specimens loaded and irradiated in material irradiation rig of Joyo can be tested in the hot-cell of MMF. In addition, the various samples could be machined from the wrapper tubes of driver and fuel test subassemblies irradiated in Joyo using by an electrical discharge machine (EDM). These samples can be characterized using the equipment such as tensile, miniature Charpy and TEM in the hot cell of MMF. In this presentation, we will introduce the details of feasible PIE equipment in the hot-cell of MMF, and examples of typical PIE data acquired by these equipment.
Tanno, Takashi; Yano, Yasuhide; Sekio, Yoshihiro; Oka, Hiroshi; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki
no journal, ,
Ferritic/martensitic (F/M) steels are candidate for core material of fast reactors (FR) because of its superior swelling resistance. A 11Cr F/M steel (PNC-FMS) have been developed for wrapper tube and cladding tube of demonstration FR in Japan Atomic Energy Agency. For demonstration of in-reactor performance and preparation of material strength standard, it is important to extend database on irradiation and thermal aging effects. In this work, ring tensile tests and hardness tests of PNC-FMS irradiated in Joyo up to 32.5 dpa at 455-835 
C were carried out, and the results were compared with those of aging tests in order to clarify the irradiation effects exclusive of thermal aging effects. The UTS at RT of PNC-FMS irradiated at over 600 C tended to be lower than those of as-received and/or thermal aged ones. The hardness showed the same trend. But, the UTS and hardness test results showed that PNC-FMS irradiated at 835 C could be harder than that of aged one.
Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki
no journal, ,
For the optimization of the fabrication process for 9Cr-ODS steels, it is important to know the influence of the processing steps, i.e. hot isostatic pressing, hot extrusion, and hot forging, on the evolution of nano-structure. In this study, the effect of manufacturing process on the evolution of the nano-structure was investigated. The investigation was performed by multi-scale method, including small angle neutron scattering. The multi-scale microstructure analysis for the sample obtained from each step of processing revealed that the hot extrusion and hot forging reduced the proportion of residual ferrite phase, resulting in the decrease of the total number density of the nano-sized particles. Moreover, the hardness decreased with decreasing the proportion of residual ferrite phase. This is clearly expressing that the residual ferrite phase is the strength controlling factor of the 9Cr-ODS steel and the control of this phase is important towards the process optimization.
Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.
no journal, ,
Research and development (R&D) program for establishing technical basis of ATFs for light water reactor (LWR) started in 2015. Since then the R&D is being conducted in cooperation with power plant providers, fuel venders, research institutes and universities for making the most use of the experiences in R&D, practical design, and evaluations of fuels and cores of commercial LWRs. Among currently explored ATF candidate materials in the program, silicon carbide composite reinforced by SiC fiber (SiC/SiC) and FeCrAl steel strengthened by dispersion of fine oxide particles (FeCrAl-ODS) offer several attractive features including the remarkable high temperature capabilities and the slow kinetics of steam oxidation reactions. This presentation will give an overview of the progress in ATF development and review the current status of data availability and integrity for the properties and behaviors of ATF candidate materials, followed by discussion on the primary differences from zirconium alloy in the behaviors in the severe accident scenarios. Finally, subjects to be solved for practical use of ATF will be summarized.
Okubo, Nariaki; Fujimura, Yuki
no journal, ,
The accelerator driven system (ADS) adopts a liquid metal of lead-bismuth eutectic (LBE) as a coolant and also spallation target to produce high energy neutrons. In this study, irradiation effect on the corrosion behavior was evaluated for the compativility of 316L stainless steel, which is candidate material of ADS target window, through the immersion test under LBE with saturated and low oxygen concentration followed by ion irradiation experiment. In the case of soaking in LBE with saturated oxygen concentration for SS316L steels at 450C, 330 hrs, non-irradiated region did not show clear oxide layer, however, irradiated region showed bi-layers of magnetite and spinel type oxides. The formation rate of oxide layer for irradiated region was about twice faster than that of non-irradiated region. This result suggests that diffusion behavior after irradiation and mass transfer in the interface between LBE and steel surface is important for understanding of irradiation effect on liquid metal corrosion.
Ebihara, Kenichi; Suzudo, Tomoaki; Yamaguchi, Masatake
no journal, ,
Phosphorus (P) atoms bring about grain boundary (GB) embrittlement in steel materials and can influence the rise of ductile-brittle transition temperature in reactor pressure vessel steels. Thus, a rate theory model for analyzing irradiation-induced GB P segregation is developed based on the atomistic processes. So far, we have incorporated the trapping process to the model based on the result of molecular dynamics (MD) simulations. However, the conventional model is used for the trapping process. In this study, we simulated the migration of a P atom in a GB. In addition, based on the consideration of the MD results, we modified the de-trapping model and applied the rate theory model to the temperature dependence of irradiation-induced GB P segregation. It was found that P atoms migrate through a gap in the GB region. In the calculated GB P segregation, the GB P coverage increased at T 600C and that the increase depended on the GB P segregation energy.