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Kikuchi, Shin; Kondo, Toshiki; Doi, Daisuke; Seino, Hiroshi; Ogawa, Kengo*; Nakagawa, Takeshi*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Morita, Koji*; Nakamura, Kinya*; Ahmed, Z.*; Pellegrini, M.*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
This paper describes the project overview and progress of experimental and analytical studies conducted until 2022. A specific result in this paper is to obtain first experimental data of BC-SS eutectic freezing.
Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
We have been developed a thermal-hydraulic analysis model in the reactor vessel using the computational fluid dynamics code with a low computational cost to evaluate core-plenum interactions during a natural circulation decay heat removal using a dipped-type direct heat exchanger in a design of sodium-cooled fast reactors. In this study, we investigate the coarse mesh modeling of interwrapper gap (IWG) using correlations for the purpose of the development of a practical model which can reduce the computational cost maintaining the prediction accuracy. An influence of combinations of the coarse mesh and the correlation for pressure loss in the IWG on the thermal-hydraulics and the core temperature distribution is revealed through the numerical analysis of a sodium experiment.
Morita, Koji*; Yamano, Hidemasa
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
This paper describes the generalized model developed for these eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as for the reactions that occur between eutectic reaction products in the solid and liquid states and SS or B
C. We also describe the thermophysical property model based on thermophysical property data.
Ahmed, Z.*; Wu, S.*; Pellegrini, M.*; Okamoto, Koji*; Sharma, A.*; Yamano, Hidemasa
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 14 Pages, 2024/08
The analysis show that once eutectic reaction occurs, the boron diffuses into the stainless steel (SS) wall. Melting initiates at the BC and SS interface, with melt flow following SS cladding penetration. Also, we observed that as temperature rises, a proportional increase in the boron concentration within the melt. The updated MPS method indicated a computational capability of the eutectic reaction model used to effectively analyze control rod eutectic reactions, simulating severe accidents, and its subsequent relocation to understand the effect of B
C ingress into the core.
Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
This paper describes the application of safety design criteria (SDC) and safety design guidelines (SDG) developed in the Generation-IV International Forum on the natural circulation of sodium to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
The VOF method is a type of CFDs that is most widely applied to multiphase flow analysis involving advective interfaces, and several interface-capturing schemes have been developed for an accurate advection of VOF values. However, the performance of these schemes has typically been evaluated only for limited numerical problems where velocity fields are spatially orderly and fixed in time. Few studies have been conducted to evaluate the performance of these schemes for more realistic and complex conditions, such as gas-liquid two-phase flows in nuclear reactors. Therefore, in this study, three-dimensional analysis of bubble flows has been conducted using the interface-capturing schemes of THINC and THINC/WLIC, which have been developed relatively recently. Evaluation is performed using more engineering indicators such as the number, volume, and trajectory of bubbles, which can influence the void fraction distribution in reactor cores. The results of these comparisons showed that the VOF value could be significantly diffused, leading to numerical brake-up and dissipation of the bubbles, with the influence of interface-capturing scheme.
Yamano, Hidemasa
no journal, ,
This reports R&D for sodium-cooled fast reactors in Japan.
Yamano, Hidemasa
no journal, ,
This reports Advanced Reactor Development in Japan, focused on SFR and HTGR.