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Journal Articles

Alloy design and characterization of a recrystallized FeCrAl-ODS cladding for accident-tolerant BWR fuels; An Overview of research activity in Japan

Ukai, Shigeharu; Sakamoto, Kan*; Otsuka, Satoshi; Yamashita, Shinichiro; Kimura, Akihiko*

Journal of Nuclear Materials, 583, p.154508_1 - 154508_24, 2023/09

 Times Cited Count:0 Percentile:95.21(Materials Science, Multidisciplinary)

Journal Articles

Development of a numerical simulation method for air cooling of fuel debris by JUPITER

Yamashita, Susumu; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 10(4), p.22-00485_1 - 22-00485_25, 2023/08

A detailed evaluation for air cooling of fuel debris in actual reactors will be essential in fuel debris retrieval under dry conditions. To understand the heat transfer in and around fuel debris, which is assumed as a porous medium in the primary containment vessel (PCV) mechanistically, we newly applied the porous medium model to the multiphase and multicomponent computational fluid dynamics code named JUPITER (JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research). We applied the Darcy-Brinkman model as for the porous medium model. This model has high compatibility with JUPITER because it can treat both a pure fluid and a porous medium phase simultaneously in the same manner as the one-fluid model in multiphase flow simulation. We addressed the case of natural convection with a high-velocity flow standing out nonlinear effects by implementing the Forchheimer model, including the term of the square of the velocity as a nonlinear effect to the momentum transport equation of JUPITER. We performed some simple verification and validation simulations, such as the natural convection simulation in a square cavity and the natural convective heat transfer experiment with the porous medium, to confirm the validity of the implemented model. We confirmed that the result of JUPITER agreed well with these simulations and experiments. In addition, as an application of the updated JUPITER, we performed the preliminary simulation of air cooling of fuel debris in the condition of the Fukushima Daiichi Nuclear Power Station unit 2 including the actual core materials. As a result, JUPITER calculated the temperature and velocity field stably in and around the fuel debris inside the PCV. Therefore, JUPITER has the potential to estimate the detailed and accurate thermal-hydraulics behaviors of fuel debris.

Journal Articles

Development of numerical simulation method of natural convection around heated porous medium by using JUPITER

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05

For contaminated water management in decommissioning Fukushima Daiichi Nuclear Power Stations, reduction in water injection, intermittent injection water and air cooling are considered. However, since there are uncertainties of fuel debris in the PCV, it is necessary to examine and evaluate optimal cooling methods according to the distribution state of the fuel debris and the progress of the fuel debris retrieval work in advance. We have developed a method for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation and the porosity of fuel debris. Since a large-scale thermal-hydraulics analysis of natural convection is necessary for the method, JUPITER developed independently by JAEA is used. It is however difficult to perform the large-scale thermal-hydraulics analysis with JUPITER by modeling the internal structure of the debris which may consist of a porous medium. Therefore, it is possible to analyze the heat transfer of the porous medium by adding porous models to JUPITER. In this study, we report the validation of JUPITER applied the porous model and discuss which heat transfer models are most effective in porous models such as series, parallel and geometric mean models. To obtain validation data of JUPITER for the natural convective heat transfer analysis around the porous medium, we performed the heat transfer and the flow visualization experiments of the natural convection in the experimental system including the porous medium. In the comparison between the experiment and the numerical analysis with each model, the numerical result with the geometric mean model was the closest of the models to the experimental results. However, the numerical results of the temperature and the velocity were overestimated for those experimental results. In particular, the temperature near the interface between the porous medium and air was more overestimated.

Journal Articles

Current status of accident tolerant fuel (ATF) development, 1; Overview of ATF development conducted under the technology development project for improving nuclear safety

Yamashita, Shinichiro

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 65(4), p.233 - 237, 2023/04

In the wake of the accident at the Fukushima Daiichi Nuclear Power Plant (NPP) of TEPCO due to the Great East Japan Earthquake in 2011, interest in the early implementation of accident tolerant fuel (ATF) not only for many existing NPPs but also for future NPPs, which is expected to dramatically improve the safety of light water reactors, has increased globally, and research and development is currently underway in many countries around the world. In this article, an overview of domestic ATF technology development that has been carried out with the support of the Ministry of Economy, Trade and Industry since 2015, will be introduced.

Journal Articles

A Numerical simulation method to evaluate heat transfer of fuel debris in air cooling by JUPITER, 1; Project overview and the applicability to the actual reactor system

Yamashita, Susumu; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10

no abstracts in English

Journal Articles

A Numerical simulation method to evaluate heat transfer of fuel debris in air cooling by JUPITER, 2; Validation of porous model for natural convective heat transfer

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10

Journal Articles

Dislocation-climbing bypass over dispersoids with different lattice misfit in creep deformation of FeCrAl oxide dispersion-strengthened alloys

Ukai, Shigeharu; Yamashita, Shinichiro

Journal of Materials Research and Technology, 16, p.891 - 898, 2022/01

 Times Cited Count:7 Percentile:87.71(Materials Science, Multidisciplinary)

The creep strain rate of FeCrAl oxide dispersion-strengthened alloys, as a promising accident-tolerant fuel (ATF) cladding of the light-water reactors, is accelerated in YAlO$$_{3}$$ dispersoids by two to three orders of magnitude compared with Y$$_{4}$$Zr$$_{3}$$O$$_{12}$$ dispersoid at 1273K and even occurs at an applied stress less than threshold stress for dislocation detachment. Two approaches were carried out to interpret new findings and to clarify their mechanism. By optimizing the relaxation of the dislocation line energy at the dispersoid interface, numerical analyses proved the accelerated dislocation-climbing in the YAlO$$_{3}$$ dispersoids. The other is a more atomistic approach. The climbing force on the dislocation induced by the stress field around the dispersoid was analyzed in terms of the Peach-Koehler relationship. The accelerated creep strain rate in YAlO$$_{3}$$ dispersoids is attributed to a larger climbing force induced by larger lattice misfit with less coherency in YAlO$$_{3}$$ dispersoid.

Journal Articles

Summary results of subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))"

Koyama, Shinichi; Nakagiri, Toshio; Osaka, Masahiko; Yoshida, Hiroyuki; Kurata, Masaki; Ikeuchi, Hirotomo; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Takano, Masahide; et al.

Hairo, Osensui Taisaku jigyo jimukyoku Homu Peji (Internet), 144 Pages, 2021/08

JAEA performed the subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))" in 2020JFY. This presentation summarized briefly the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning and Contaminated Water Management.

Journal Articles

Orientation dependence of yield strength in a new single crystal-like FeCrAl oxide dispersion strengthened alloy

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$$^{circ}$$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.

Journal Articles

Locally mesh-refined lattice Boltzmann method for fuel debris air cooling analysis on GPU supercomputer

Onodera, Naoyuki; Idomura, Yasuhiro; Uesawa, Shinichiro; Yamashita, Susumu; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00531_1 - 19-00531_10, 2020/06

A dry method is one of practical methods for decommissioning the TEPCO's Fukushima Daiichi Nuclear Power Station. Japan Atomic Energy Agency (JAEA) has been evaluating the air cooling performance of the fuel debris by using the JUPITER code based on an incompressible fluid model and the CityLBM code based on the lattice Boltzmann method (LBM). However, these codes were based on a uniform Cartesian grid system, and required large computational time and cost to capture complicated debris structures. We develop an adaptive mesh refinement (AMR) version of the CityLBM code on GPU based supercomputers and apply it to thermal-hydrodynamics problems. The proposed method is validated against free convective heat transfer experiments at JAEA. It is also shown that the AMR based CityLBM code on 4 NVIDIA TESLA V100GPUs gives 6.7x speedup of the time to solution compared with the JUPITER code on 36 Intel Xeon E5-2680v3 CPUs.

Journal Articles

Microstructure and texture evolution and ring-tensile properties of recrystallized FeCrAl ODS cladding tubes

Aghamiri, S. M. S.*; Sowa, Takashi*; Ukai, Shigeharu*; Ono, Naoko*; Sakamoto, Kan*; Yamashita, Shinichiro

Materials Science & Engineering A, 771, p.138636_1 - 138636_12, 2020/01

 Times Cited Count:25 Percentile:90.87(Nanoscience & Nanotechnology)

Oxide dispersion strengthened (ODS) FeCrAl ferritic steels are being developed as potential accident tolerance fuel cladding materials for the light water reactors (LWRs) due to significant improvement in steam oxidation by alumina forming scale and good mechanical properties up to high temperatures. In this study, the microstructural characteristics and tensile properties of the two FeCrAl ODS cladding tubes with different extrusion temperatures of 1100$$^{circ}$$C and 1150$$^{circ}$$C were investigated during processing conditions. While the hot extruded sample showed micron sized elongated grains with strong $$alpha$$-fiber in $$<$$110$$>$$ texture, cold pilger rolling process change the microstructure to submicron/micron size grain structure along with texture evolution to both $$alpha$$-fiber ($$<$$110$$>$$ texture) and $$gamma$$-fiber ({111} texture) via crystalline rotations. Subsequently, final annealing resulted in evolution of microstructure to large grain recrystallized structure starting at recrystallization temperature of $$sim$$810-850$$^{circ}$$C. Two distinct texture development happened in recrystallized cladding tubes, i.e., only large elongated grains of (110) $$<$$211$$>$$ texture following extrusion temperature of 1100$$^{circ}$$C; and two texture components of (110) $$<$$211$$>$$ and {111} $$<$$112$$>$$ following higher extrusion temperature of 1150$$^{circ}$$C. The different texture development and retarding of recrystallization progress in 1100$$^{circ}$$C-extruded cladding tubes were attributed to higher distribution of oxide particles.

Journal Articles

Oxidation of silicon carbide in steam studied by laser heating

Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Furumoto, Kenichiro*; Sato, Hisaki*; Ishibashi, Ryo*; Yamashita, Shinichiro

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.670 - 674, 2019/09

Journal Articles

Overview of accident-tolerant fuel R&D program in Japan

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.

Journal Articles

Fuel debris' air cooling analysis using a lattice Boltzmann method

Onodera, Naoyuki; Idomura, Yasuhiro; Kawamura, Takuma; Uesawa, Shinichiro; Yamashita, Susumu; Yoshida, Hiroyuki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 6 Pages, 2019/05

A dry method is one of practical methods for decommissioning the TEPCO's Fukushima Daiichi Nuclear Power Station. Japan Atomic Energy Agency (JAEA) has been evaluating the air cooling performance by using the JUPITER code. However, the JUPITER code requires a large computational cost to capture debris' structures. To accelerate such CFD analyses, we use the CityLBM code, which is based on the lattice Boltzmann method (LBM) and is highly optimized for GPUs. The CityLBM code is validated against free convective heat transfer experiments at JAEA, and the similar accuracy as the JUPITER code is confirmed regarding the prediction capability of heat transfer and the resulting temperature distributions. It is also shown that the elapse time of a CityLBM simulation on GPUs is reduced to 1/6 compared with that of the corresponding JUPITER simulation on CPUs with the same number of GPUs and CPUs. The results show that the LBM is promising for accelerating thermal convective simulations.

Journal Articles

Free convective heat transfer experiment to validate air-cooling performance analysis of fuel debris

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

Journal Articles

Corrosion behaviour of FeCrAl-ODS steels in nitric acid solutions with several temperatures

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 60$$^{circ}$$C, 80$$^{circ}$$C 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.

Journal Articles

Validation of free-convective heat transfer analysis with JUPITER to evaluate air-cooling performance of fuel debris in dry method

Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 5(4), p.18-00115_1 - 18-00115_13, 2018/08

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:15 Percentile:85.51(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

Prediction of chemical effects of Mo and B on the Cs chemisorption onto stainless steel

Di Lemma, F. G.; Yamashita, Shinichiro; Miwa, Shuhei; Nakajima, Kunihisa; Osaka, Masahiko

Energy Procedia, 127, p.29 - 34, 2017/09

 Times Cited Count:4 Percentile:90.94

Chemical effects of molybdenum (Mo) and boron (B), which were considered to form compounds with Cs, on the Cs chemisorption were predicted using a chemical equilibrium calculation. It is seen that Cs$$_{2}$$MoO$$_{4}$$ were formed in the chemisorbed compounds. On the other hand, little effects were observed for B. The results suggest that the effects of Mo should be considered for further experimental investigation.

Journal Articles

Fuel behavior analysis for accident tolerant fuel with sic cladding using adapted FEMAXI-7 code

Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09

Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 9$$times$$9 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.

249 (Records 1-20 displayed on this page)