Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Ezure, Toshiki; Akimoto, Yuta; Onojima, Takamitsu; Kurihara, Akikazu; Tanaka, Masaaki
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.3652 - 3662, 2023/08
In order to grasp the thermal-hydraulic behaviors during decay heat removal by dipped-direct heat exchangers (D-DHXs) in a sodium-cooled fast reactor, an experimental study was performed using a sodium experimental facility. The simulated core of PLANDTL-2 was formed by 55 hexagonal-shaped flow channel tubes, which allows to examine the cooling behavior of the simulated core region having multiple rows of fuel assemblies including the thermal hydraulic behavior to the radial direction. In this study, transient core cooling behavior in the situation after the scram with the decay heat removal using a D-DHX was examined. The time evolution of the temperature was measured in the whole system including the simulated core region. As the results, it was confirmed there was not excessively skewed temperature distribution in the radial direction in the core region.
Yamano, Hidemasa; Morita, Koji*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4295 - 4308, 2023/08
This study applied the SIMMER-IV code with the newly developed model to a preliminary SA analysis of the SFR. The analysis results show that the eutectic reaction is caused by the contact between the liquid SS and the broken B
C pellets which are released to the coolant channel after the failure of cladding which is melted by the mixture of liquid SS and fuel particles coming from the neighboring fuel assemblies. The liquid eutectic material formed by the reaction moves from the control assembly to the neighboring fuel assemblies. The lower density of the eutectic melt than molten SS drives the upward motion of the eutectic in the molten core pool. This analysis indicated that the SIMMER-IV code using the eutectic reaction model has successfully simulated the eutectic reaction and the relocation of the eutectic melt as well as the reactivity transient behavior caused by the molten core material relocation.
Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi; Yanagisawa, Hideki*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08
The leakage of pressurized water from a steam generator (SG) and the progress after that are a key issue in the safety assessment or design of a SG in sodium-cooled fast reactor. The analysis code LEAP-III can evaluate a rate of water leakage during the long-term event progress, i.e., from the self-wastage initiated by an occurrence of a microscopic crack in a tube wall to the water leak detection and water/water-vapor blowdown. Since LEAP-III consists of semi-empirical formulae and one-dimensional equations of conservation, it has an advantage in short computation time. Thus, LEAP-III can facilitate the exploration of various new SG designs in the development of innovative reactors. However, there are several problems, such as an excessive conservative result in some case and the need for numerous experiments or preliminary analyses to determine tuning parameters of models in LEAP-III. Hence, we have developed a Lagrangian particle method code, which is characterized by a simpler computational principle and faster calculation. In this study, we have improved the existing particle pair search method for interparticle interaction in this code and developed an alternative model without the pair search. Through the trial analysis simulating in a tube bundle system, it was confirmed that new models reduced the computation time. In addition, it was shown that representative temperatures of the heat-transfer tubes evaluated by this particle method code, which is used to predict the tube failure in LEAP-III, were good agreement with that by SERAPHIM, which is a detailed mechanistic analysis method code.
Ariyoshi, Gen; Saruta, Koichi; Kogawa, Hiroyuki; Futakawa, Masatoshi; Maeno, Koki*; Li, Y.*; Tsutsui, Kihei*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1407 - 1420, 2023/08
Cavitation damage on a target vessel due to proton beam-induced pressure waves is one of the crucial issues for the pulsed neutron source using a mercury spallation target. As a mitigation technique for the damage, the helium microbubble injection into the mercury has been carried out by using a swirl bubbler in order to utilize compressibility of bubbles. Moreover, double-walled structure, which consists of an outer wall and an inner wall, has been applied as the target head structure. In this study, we aim to develop an abnormality diagnostic technology to detect the inner wall cracking, which is caused by such cavitation damage, from the outside of the target vessel. The mercury flow fields in the case with the cracking are evaluated by computational fluid dynamics analysis based on finite element method. And then, effect of the cracking on the flow field is discussed from the point of view of the flow-induced vibration and the acoustic vibration.
Maruyama, Yu; Sugiyama, Tomoyuki*; Shimada, Asako; Lind, T.*; Bentaib, A.*; Sogalla, M.*; Pellegrini, M.*; Albright, L.*; Clayton, D.*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4782 - 4795, 2023/08
Lind, T.*; Kalilainen, J.*; Marchetto, C.*; Beck, S.*; Nakamura, Koichi*; Kino, Chiaki*; Maruyama, Yu; Kido, Kentaro; Kim, S. I.*; Lee, Y.*; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4796 - 4809, 2023/08
Watanabe, Nao; Yamashita, Susumu; Uesawa, Shinichiro; Nishihara, Kenji; Yoshida, Hiroyuki
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.3522 - 3534, 2023/08
Accelerator-driven system (ADS), the coolant of which is lead-bismuth eutectic (LBE), has been designed by Japan Atomic Energy Agency. Estimating corrosion rate at the wall surface of LBE channel is an important issue in considering safety and the life of the entire structure. The corrosion rate depends on state of oxygen layers forming at the material surface. Therefore, this study aims to develop a method to evaluate the corrosion rate in ADS for the design study by estimation of the oxide layer growth and dissolution (OLGD) rates by means of numerical analysis. The OLGD rates, mass transfer rates of oxygen and iron between the material and LBE and advection-diffusion rates of them in LBE depend on each other. Therefore, in order to estimate OLGD rates, the three numerical analysis models should be coupled. For the advection-diffusion calculation, to use CFD code should be reasonable approach to analyze complex flow in ADS, while for the OLGD and the mass transfer calculation, to use some correlation equations should be reasonable because their scales are much smaller than the advection-diffusion. The present work has developed the analysis method of OLGD rates by using JUPITER code, which is CFD code developed in JAEA. In terms of the correlation equations of OLGD and mass transfer rates, existing models used in a previous study were used with modified.
Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Alzahrani, H.*; Sakai, Takaaki*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1262 - 1275, 2023/08
Development of evaluation method for cover gas entrainment by vortices generated at free surface in upper plenum of sodium-cooled fast reactor is required, and an evaluation method by predicting vortices from flow velocity distribution obtained by CFD analysis is developed. In this study, Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis. Initial mesh was refined with two indexes: the first index (Index-1) is when the second invariant, Q, of velocity gradient tensor is negative and the second one (Index-2) is pressure gradient index added to Index-1. As a result of applying AMR method to unsteady vortices system with a flat plate and performing transient analyses with refined meshes, the result of pressure distribution and velocity around the flat plate in mesh using Index-2 was similar to the result of all refined mesh. It was also confirmed that vortices generation and growth was better simulated by refining meshes around separation area.
Hasegawa, Yuta
no journal, ,
We explain the research plan for the exascale computational fluid dynamics (CFD) in Japan Atomic Energy Agency, Center for Computational Science and e-Systems (JAEA-CCSE). JAEA-CCSE has been promoting the development of 1m scale real-time urban wind simulation code, "CityLBM." In this presentation, we show our past studies that accelerate the CFD code, and our recent research which introduces ensemble data assimilation toward the digital twin of the urban wind. The outlook for the exascale computing with them is also described.
