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Kikuchi, Shin; Takai, Toshihide; Yamano, Hidemasa; Sakamoto, Kan*
Proceedings of 28th International Conference on Nuclear Engineering; Nuclear Energy the Future Zero Carbon Power (ICONE 28) (Internet), 9 Pages, 2021/08
In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B
C) and stainless steel (SS) may occur. Thus, behavior of BC-SS eutectic melting is one of the phenomena to evaluate the core disruptive accidents in SFR. In this study, the reaction experiments using SS crucibles and the pellets of SS with low BC concentration as samples were performed to simulate the state of the reaction interface in which the eutectic reaction and interdiffusion of BC-SS have progressed to a certain extent. It was revealed that the rate constants of eutectic reaction between SS and SS with low BC concentration are smaller than that of BC-SS eutectic reaction at high temperatures.
Aghamiri, S. M. S.*; Sugawara, Naoya*; Ukai, Shigeharu; Ono, Naoko*; Sakamoto, Kan*; Yamashita, Shinichiro
Materials Characterization, 176, p.111043_1 - 111043_6, 2021/06
Advanced oxidation-resistant FeCrAl ODS alloys were developed via the control of composition-processing conditions for the accident tolerant fuel (ATF) cladding. For the first time, a single-crystal like recrystallized FeCrAl ODS alloy was achieved with a unique crystallographic texture of 110-plane and 211-direction and a high number density of fine nanoscale oxide particles. Evaluation of yield strengths at different temperatures showed higher values in transverse (T) direction than longitudinal (L) direction. The crystal orientation dependence of the yield strength up to 800
C was attributed to lower value of Schmid factor in transverse direction. Accordingly, the critical resolved shear stress of this practical class of advanced materials was calculated in various temperatures.
Kikuchi, Shin; Sakamoto, Kan*; Takai, Toshihide; Yamano, Hidemasa
Nihon Kikai Gakkai 2020-Nendo Nenji Taikai Koen Rombunshu (Internet), 4 Pages, 2020/09
In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (BC) as control rod element and stainless steel (SS) as control rod cladding or related structure may occur. Thus, behavior of BC-SS eutectic melting is one of the phenomena to evaluate the core disruptive accidents in SFR. In order to clarify the kinetic feature of BC-SS eutectic melting process in the interface, the thinning test for SS crucibles using the pellets of BC or SS with low BC concentration were performed to obtain the rate constant with dependence of BC concentration against SS. It was found that the rate constants of eutectic melting between SS and SS low BC concentration were smaller than that of BC-SS in the high temperature range. Besides, the rate constant of eutectic melting between SS and BC containing SS became small when decreasing the BC concentration against SS.
Kim, J.*; Yamanaka, Satoru*; Murayama, Ichiro*; Kato, Takanori*; Sakamoto, Tomokazu*; Kawasaki, Takuro; Fukuda, Tatsuo; Sekino, Toru*; Nakayama, Tadachika*; Takeda, Masatoshi*; et al.
Sustainable Energy & Fuels (Internet), 4(3), p.1143 - 1149, 2020/03
Times Cited Count：3 Percentile：51.8(Chemistry, Physical)Ohashi, Kunihide*; Hino, Takanori*; Kobayashi, Hiroshi*; Onodera, Naoyuki; Sakamoto, Nobuaki*
Journal of Marine Science and Technology, 24(3), p.884 - 901, 2019/09
Times Cited Count：5 Percentile：58.68(Engineering, Marine)An unsteady Reynolds averaged Navier-Stokes solver with a structured overset grid method has been developed. Velocity pressure coupling is achieved using an artificial compressibility approach, spatial discretization is based on a FVM. Body motions are considered using the grid deformation technique and grid velocities in the convective term. The full multigrid (FMG) method is applied to obtain fast convergence. The cell flag on a coarse grid level is determined using the cell flag on a fine grid level. In the coarse and fine grid level calculations at the FMG stage, the data are interpolated until the finest grid level is achieved at an overset update interval. Then, the data are updated based on the overset relations at the finest grid level and then transferred to a coarser grid level. The computations for flows around a hull form, including an unsteady simulation with regular waves, are demonstrated.
Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09
After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.
Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05
ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.
Sakamoto, Tetsuo*; Morita, Masato*; Kanenari, Keita*; Tomita, Hideki*; Sonnenschein, V.*; Saito, Kosuke*; Ohashi, Masaya*; Kato, Kotaro*; Iguchi, Tetsuo*; Kawai, Toshihide*; et al.
Analytical Sciences, 34(11), p.1265 - 1270, 2018/11
Times Cited Count：7 Percentile：39.71(Chemistry, Analytical)Sano, Yuichi; Sakamoto, Atsushi; Takeuchi, Masayuki; Misumi, Ryuta*; Kunii, Kanako*; Todoroki, Kei*; Nishi, Kazuhiko*; Kaminoyama, Meguru*
Kagaku Kogaku Rombunshu, 44(6), p.335 - 340, 2018/11
Concerning an annular centrifugal contactor which has high throughput and separation performance, the effect of operational condition on fluidic and dispersion behavior, which are important to improve the contactor performance, was investigated by computational fluid dynamics (CFD) analysis based on the turbulence model, and the calculated results were validated by experimental data. The liquid phase in the annular zone was gradually divided into two regions vertically with increasing the rotor speed and decreasing the flowrate, and the liquid flow moved toward the center of the housing bottom was generated in the lower annular zone under any operational condition. The droplet size of the dispersed phase in the annular zone decreased with increasing the rotor speed and decreasing the flowrate. These calculation results showed a good agreement with experimental data. The CFD analysis considering mass transfer between aqueous and organic phases was also attempted, and it was confirmed that the change of extraction performance with the rotor speed showed the same tendency as the experimental result.
Co
)Fe
O (x=0-1) layers on Si(111) studied by soft X-ray magnetic circular dichroismWakabayashi, Yuki*; Nonaka, Yosuke*; Takeda, Yukiharu; Sakamoto, Shoya*; Ikeda, Keisuke*; Chi, Z.*; Shibata, Goro*; Tanaka, Arata*; Saito, Yuji; Yamagami, Hiroshi; et al.
Physical Review Materials (Internet), 2(10), p.104416_1 - 104416_12, 2018/10
Times Cited Count：6 Percentile：37.83(Materials Science, Multidisciplinary)Yamazaki, Saishun; Pshenichnikov, A.; Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Tokushima, Kazuyuki*; Aomi, Masaki*; Sakamoto, Kan*
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 8 Pages, 2018/10
It is predictively evaluated that degradation of fuel assembly proceeded in a certain steam-starved condition at the early stage of a SA at 1F unit 2 (BWR). As for PWR fuel assembly, effective steam flow rate was properly indicated by normalizing to a unit of g-HO/sec/rod which is used as an important parameter for evaluating fuel degradation progression. Due to the inhomogeneous configuration of BWR fuel assembly, the difference of Zry oxidation and hydrogen uptake between the inside and outside of the channel box cannot be properly evaluated by this normalization. Instead of g-HO/sec/rod, proper evaluation unit for BWR configuration is necessary. To accumulate Zry oxidation and hydrogen uptake data for steam-starved conditions, high temperature oxidation tests were performed using a simulated BWR fuel bundle sample. The use of equivalent diameter of the cross section of BWR fuel assembly was proposed for normalization of effective steam flow rate.
Takahatake, Yoko; Ambai, Hiromu; Sano, Yuichi; Takeuchi, Masayuki; Koizumi, Kenji; Sakamoto, Kan*; Yamashita, Shinichiro
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 9 Pages, 2018/10
The corrosion behaviour of FeCrAl-ODS steels for the accident tolerant fuel cladding of LWRs were investigated in nitric acid solutions for the reprocessing process of spent fuels. The corrosion tests were carried out at 60C, 80C and the boiling point of the solutions, and the specimens were then analysed by XPS. The corrosion remarkably progressed at the boiling point, and the highest corrosion rate was 0.22 mm/y. In the oxide film, the atomic concentration of Fe was lower, than that in the base material, and those of Cr and Al were higher. The results show that the corrosion of FeCrAl-ODS steels in hot nitric acid solution is not severe because of the high corrosion resistance of the oxide film formed on the material; hence, the corrosion resistance of the new cladding materials in the dissolution process of spent fuel is acceptable for reprocessing operations.
Misumi, Ryuta*; Todoroki, Kei*; Kunii, Kanako*; Nishi, Kazuhiko*; Kaminoyama, Meguru*; Sano, Yuichi; Sakamoto, Atsushi; Takeuchi, Masayuki
Kagaku Kogaku Rombunshu, 44(5), p.285 - 291, 2018/09
Annular centrifugal extractors have been anticipated for use as extractors in spent nuclear fuel recycling. The extraction rate and the liquid-liquid dispersion are related to the flow pattern in the vessel. However, no study has clarified flow patterns in vessels of various scales. For this study, flow pattern characteristics are quantified for extractors of two scales. An extractor has a mixing zone around the vessel bottom and a separation zone in the cylindrical rotor. For this experiment, distilled water was fed into the vessel. Flow behavior in the mixing zone was observed from a side view using a digital video camera at various rotor speeds and supply flow rates for extractors of two scales. In some cases, the liquid horizontal velocity vectors in the mixing zone were measured using particle image velocimetry. Results demonstrate that flow behaviors in the mixing zone in both scales of extractors are classifiable as three types, changing with operational conditions: Type A, Type B, and a Transition regime. For the Type A state, the mixing zone is fully filled with liquid from the vessel bottom up to the lower edge of the rotor. In the Type B state, the zone with existing liquid is vertically divisible into two regions. Lower rotor speeds and higher flow rates tend to produce Type A state flow behavior. The boundary operational condition between Type A and the Transition regime are correlated with the normalized supply flow rate and pumping capacity of the rotor, which is evaluated from liquid surface level in a rotor formed by centrifugal force. Furthermore, the fluid velocity in the mixing zone is roughly proportional to the rotor surface circumferential speed irrespective of the vessel scale.
Misumi, Ryuta*; Kunii, Kanako*; Todoroki, Kei*; Nishi, Kazuhiko*; Kaminoyama, Meguru*; Sano, Yuichi; Sakamoto, Atsushi; Takeuchi, Masayuki
Kagaku Kogaku Rombunshu, 44(3), p.135 - 141, 2018/05
Times Cited Count：0 Percentile：0(Engineering, Chemical)Annular centrifugal extractors have been used in spent nuclear fuel reprocessing, but the relation between the extraction rate and flow pattern in the vessel remains unclear. This study quantifies characteristics of the flow pattern to clarify this relation. An extractor produces a mixing zone around the vessel bottom and a separation zone in the rotor. The horizontal velocity of the liquid in the mixing zone was measured using particle image velocimetry at various rotor speeds and supply flow rates. Flow behaviors in the mixing zone are of three types, changing with operational conditions: Type A, Type B, and a transition regime. At lower rotor speeds and high supply flow rates, the mixing zone is fully filled with liquid from the vessel bottom up to the lower edge of the rotor: the Type A flow state. At high rotor speeds and low supply flow rates, the zone with existing liquid is vertically divisible into two regions: near the vanes and around the bottom of the rotor, which is the Type B flow state. A transition regime is also observed between Type A and Type B state. In each region surrounding the two vanes on the vessel bottom and the vessel wall, the liquid flowed in the direction of rotor rotation along the vessel wall. Liquid flow altered by the vane flowed toward the center of vessel bottom. The liquid then entered the separation zone through the orifice at the rotor bottom. For the Type A state, the horizontal velocity distribution was roughly proportional to the rotor speed. For the Type B state, the horizontal velocities around the vessel bottom were lower than those of Type A and were not proportional to the rotor speed. Presumably, the liquid fed into the vessel went directly to the rotor instead of passing between the two vanes attached to the vessel bottom.
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：6 Percentile：64.98(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 300C. 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 
' phase transformation.
Muta, Hiroaki*; Nishikane, Ryoji*; Ando, Yusuke*; Matsunaga, Junji*; Sakamoto, Kan*; Harjo, S.; Kawasaki, Takuro; Oishi, Yuji*; Kurosaki, Ken*; Yamanaka, Shinsuke*
Journal of Nuclear Materials, 500, p.145 - 152, 2018/03
Times Cited Count：4 Percentile：56.63(Materials Science, Multidisciplinary)Theis, C.*; Carbonez, P.*; Feldbaumer, E.*; Forkel-Wirth, D.*; Jaegerhofer, L.*; Pangallo, M.*; Perrin, D.*; Urscheler, C.*; Roesler, S.*; Vincke, H.*; et al.
EPJ Web of Conferences, 153, p.08018_1 - 08018_5, 2017/09
Times Cited Count：0 Percentile：0.03At CERN, gas-filled ionization chambers PTW-34031 (PMI) are commonly used in radiation fields including neutrons, protons and 
-rays. A response function for each particle is calculated by the radiation transport code FLUKA. To validate a response function to high energy neutrons, benchmark experiments with quasi mono-energetic neutrons have been carried out at RCNP, Osaka University. For neutron irradiation with energies below 200 MeV, very good agreement was found comparing the FLUKA simulations and the measurements. In addition it was found that at proton energies of 250 and 392 MeV, results calculated with neutron sources underestimate the experimental data due to a non-negligible gamma component originating from the target 
Li(p,n)Be reaction.
Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.
Takano, Sho*; Kusagaya, Kazuyuki*; Goto, Daisuke*; Sakamoto, Kan*; Yamashita, Shinichiro
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
We focused on one of accident tolerant fuel (ATF) materials, Oxide Dispersion Strengthened Fe-Cr-Al Steel (FeCrAl-ODS). There is a reasonable prospect that FeCrAl-ODS is applied to BWRs, but relatively high neutron absorption should be compensated. To decrease adverse neutron economic impact, thin FeCrAl-ODS cladding was designed, and we evaluated characteristics of a core into which 9
9 Advanced BWR (ABWR) bundles with thin FeCrAl-ODS claddings were loaded. Thin FeCrAl-ODS water rods and channel boxes were also applied. We confirmed that FeCrAl-ODS core reactivity was sufficient by increasing enrichment of UO fuel under the limit of 5 wt%. Moreover, some representative FeCrAl-ODS core characteristics were comparable to zircaloy core. We also confirmed that fuel thermal-mechanical behaviors of thin FeCrAl-ODS cladding at normal operation and transient conditions were acceptable. These results led to a conclusion that FeCrAl-ODS was applicable to BWR in the analysis range of this study.
Kimura, Akihiko*; Yuzawa, Sho*; Sakamoto, Kan*; Hirai, Mutsumi*; Kusagaya, Kazuyuki*; Yamashita, Shinichiro
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
The effect of Al addition on the PRW weldability of ODS steel is shown with the discussion focusing on the microstructure changes by the welding. The ordinary welding methods including electron beam (EB) welding and tungsten inert gas (TIG) welding were also applied to the SUS430 endcap welding to cladding tube made of FeCrAl-ODS steel. The endcap welded ODS steel tube samples were tensile tested at RT. The EB welded FeCrAl-ODS/SUS430 samples broke in the ODS steel tube, indicating that the weld bond is stronger than the ODS base metal. However, the TIG welded FeCrAl-ODS/SUS430 samples broke at a weld bond. X-ray CT scan analysis was performed for the weld bond, and the bonding strength was correlated with the X-ray CT results in order to assess the feasibility of those welding methods for ATF-ODS steel cladding.
