Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Kamiya, Tomohiro; Yoshida, Hiroyuki
Proceedings of the Symposium on Shock Waves in Japan (Internet), 7 Pages, 2024/03
We developed a ghost fluid method satisfying conservation laws to simulate steam explosions that can occur at the accident of a nuclear power plant. In the developed method, a first-order approximation is applied to interface effect regions, and a high-order approximation is applied to bulk regions. In other words, the algorithm of the developed method is not consistent. Therefore, we modify the way of getting ghost fluids and propose a comprehensive algorithm that applies a high-order approximation to interface effect regions. In the presentation, we will report the outlines and results of the numerical tests of it.
Tanaka, Masaaki; Miyake, Yasuhiro*; Ezure, Toshiki; Hamase, Erina
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The numerical analysis model for the computational fluid dynamics (CFD) code for the design study is developed to evaluate the thermal-hydraulics in the core under the core-plenum interaction (CPI) during the decay heat removal using the dipped type direct heat exchanger (D-DHX). To judge the adequacy of the numerical results for a validation study with the sodium experiment results conducted at PLANDTL-2 facility, the degree of difference (DoD) between the numerical and experimental results must be measured by using the area validation metrics (AVM). Through the examinations, the applicability of the AVM and MAVM based on the p-box method was confirmed.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Kam, D. H.*; Grabaskas, D.*; Starkus, T.*; Bucknor, M.*; Uchibori, Akihiro
Transactions of the American Nuclear Society, 126(1), p.536 - 539, 2022/06
Removal of gaseous radionuclides from the bubbles released into the sodium pool is an important consideration of fuel pin failure accident in sodium-cooled fast reactors. To support modeling of this phenomenon as a part of development of the SRT (Simplified Radionuclide Transport) code in Argonne National Laboratory, numerical analysis of experiment on Iodine gas transport to sodium pool was performed. A proposed evaluation method can be regarded to be reasonably predicting the measured decontamination factors.
Uchibori, Akihiro; Sonehara, Masateru; Aoyagi, Mitsuhiro; Takata, Takashi*; Ohshima, Hiroyuki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
A new computational code, SPECTRA, has been developed for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors. The in-vessel thermal hydraulics module includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.
Kitagawa, Yoshihiro; Shirahama, Takuma*; Kisohara, Naoyuki; Tsuboi, Akihiko
Dai-96-Kai Reza Kako Gakkai Koen Rombunshu (Internet), p.91 - 96, 2022/01
Laser scanning quenching is a locally and rapidly heat-treated process and has an advantage of no coolant required. Compared with conventional technique such as induction quenching, the region of laser quenching is about 0.5
0.7mm in depth and it needs to be expanded for more applications or durability. For this purpose, the temperature distributions and transitions in materials during laser irradiation have been revealed by using a 3D heat transfer computer code, micro-structural observation and hardness transitions in depth direction. The results indicate the laser irradiation with low power and low scan speed condition allows deeper quenching area, but it also suggests the hardness of the deepest quenching area is degraded due to slow temperature decreasing rate after laser heat scanning. Multiple times continuous irradiation have been proposed and studied to resolve this hardness degradation, and maximum quenching depth of 1.4mm is obtained under three times irradiation and controlling its power and scan speed properly.
Madokoro, Hiroshi; Sato, Ikken
Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05
Times Cited Count：6 Percentile：72.21(Nuclear Science & Technology)Yamamoto, Masahiro
Keisan Kogaku, 25(3), p.4105 - 4108, 2020/07
Recently, some attempts using Multi-Physics simulation for corrosion problems, especially crevice corrosion, have been increasing. Corrosion undergoes by electrochemical reaction. The numerical calculation procedure is used a non-liner equation. Furthermore, this reaction is affected by environmental factors, i.e. composition, amount and mobility of chemical species and redox potential. These values change with time by corrosion process itself. This report, these needs for Multi-Physics calculations are introduced.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Muramatsu, Toshiharu; Sato, Yuji; Kamei, Naomitsu; Aoyagi, Yuji*; Shobu, Takahisa
Nihon Kikai Gakkai Dai-13-Kai Seisan Kako, Kosaku Kikai Bumon Koenkai Koen Rombunshu (No.19-307) (Internet), p.157 - 160, 2019/10
no abstracts in English
Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08
Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Uesawa, Shinichiro; Suzuki, Takayuki*; Yoshida, Hiroyuki
Konsoryu Shimpojiumu 2018 Koen Rombunshu (Internet), 2 Pages, 2018/08
no abstracts in English
Hirakawa, Moe*; Kikuchi, Yuichiro*; Sakai, Takaaki*; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Gas entrainment (GE) from cover gas is one of key issue for Sodium-cooled fast reactors to prevent unexpected effects to core reactivity. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments which were generated as wake vortexes behind a plate obstacle in the circulating water channel. In this paper, the results of comparison between experiments and analyses were discussed and the gas core lengths from the surface vortexes were evaluated by using the evaluation tool named StreamViewer developed by Japan Atomic Energy Agency.
Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
no abstracts in English
Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.
Muramatsu, Toshiharu
Hikari Araiansu, 28(12), p.31 - 35, 2017/12
no abstracts in English
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08
A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.
Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*
Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2017/06
In JAEA, a numerical simulation code named MUGTHES which can deal with conjugate heat transfer between the fluid and the structure parts has been developed for estimation of the thermal fatigue issue. In fundamental validation, the benchmark analysis was considered using the experiment of planar triple parallel jet sodium test (PLAJEST). Three specific experimental conditions at Vr=1, 1.56, and 5.56 were employed for the benchmark analyses according to the knowledge in the literatures. Through the benchmarks, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model in MUGTHES to simulate thermal striping phenomena was potentially confirmed and issues to be modified in the future works were indicated.
Muramatsu, Toshiharu
Reza Kenkyu, 44(12), p.799 - 803, 2016/12
no abstracts in English
Horiguchi, Naoki; Yoshida, Hiroyuki; Nakao, Yasuhiro*; Kaneko, Akiko*; Abe, Yutaka*
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
