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C-stainless steel using the Moving Particle Semi-Implicit (MPS) MethodAhmed, 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 BC ingress into the core.
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 BC. We also describe the thermophysical property model based on thermophysical property data.
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.
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.
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.
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
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
Matsushita, Hatsuki*; Kobayashi, Ren*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 9 Pages, 2022/09
During core disruptive accidents in sodium-cooled fast reactors, the molten core material flows through flow channels, such as the control rod guide tubes, into the core inlet plenum under the core region. The molten core material can be cooled and solidified while impinging on a horizontal plate of the inlet plenum in a sodium coolant. However, the solidification and cooling behaviors of molten core materials impinged on a horizontal structure have not been sufficiently studied thus far. Notably, this is an important phenomenon that needs to be elucidated from the perspective of improving the safety of sodium-cooled fast reactors. Accordingly, a series of experiments on discharging a simulated molten core material (alumina: Al
O
) into a sodium coolant on a horizontal structure was conducted at the experimental facility of the National Nuclear Center of the Republic of Kazakhstan. In this study, analyses on the sodium experiments using SIMMER-III as the fast reactor safety evaluation code were performed. The analysis methods were validated by comparing the results and experiment data. In addition, the cooling and solidification behaviors during jet impingement were evaluated. The results indicated that the molten core material exhibited fragmentation owing to the impingement on the horizontal plate and was, therefore, scattered toward the periphery. Furthermore, the simulated molten core material was evaluated to be cooled by sodium and subsequently solidified.
Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
To enhance the safety of sodium-cooled fast reactors, the natural circulation (NC) decay heat removal systems with a dipped-type direct heat exchanger (D-DHX) have been investigated. During the D-DHX operation, since the core-plenum interaction occurs, development of the reactor vessel model including the more model by using a computational fluid dynamics code (RV-CFD) is required. Previously, the CFD model based on the subchannel analysis was developed. In this study, to achieve much lower computational cost maintaining the prediction accuracy, the coarse-mesh subchannel CFD (CMSC) model has been developed and was incorporated into the core of RV-CFD. As a result of PLANDTL-1 test analysis, the RV-CFD with the CMSC model can reproduce the radial heat transfer under NC conditions.
Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Morita, Koji*; Nakamura, Kinya*; Fukai, Hirofumi*; et al.
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
This paper describes the project overview and progress of experimental and analytical studies conducted until 2020. Specific results in this paper are the measurement of the eutectic reaction rates and the validation of physical model describing the eutectic reaction in the analysis code through the numerical analysis of the BC-SS eutectic reaction rate experiments in which a BC pellet was placed in a SS crucible.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.
Ono, Ayako; Yamashita, Susumu; Sakashita, Hiroto*; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
Japan Atomic Energy Agency is developing the computational fluid dynamics code, JUPITER, based on the volume of fluid (VOF) method to analyze detailed thermal-hydraulics in a reactor. The detailed numerical simulation of boiling from a heating surface needs a substantial computational cost to resolve the microscale thermal-hydraulic phenomena such as the bubble generation from a cavity and evaporation of a micro-layer. This study developed the simplified boiling model from the heating surface to reduce the computational cost, which will apply to the detailed simulation code based on the surface tracking method such as JUPITER. We applied the simplified boiling model to JUPITER, and compared the simulation results with the experimental data of the vertical heating surface in the forced convection. We confirmed the degree of their reproducibility, and the issues to be modified were extracted.
Phan, L. H. S.*; Ohara, Yohei*; Kawata, Ryo*; Liu, X.*; Liu, W.*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Tagami, Hirotaka
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
Self-leveling behavior of core fuel debris beds is one of the key phenomena for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The SIMMER code has been developed for CDA analysis of SFRs, and the code has been successfully applied to numerical simulations for key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, in SIMMER's fluid-dynamics model, it is always difficult to represent the strong interactions between solid particles as well as the discrete particle characteristics. To solve this problem, a new method has been developed by combining the multi-fluid model of the SIMMER code with the discrete element method (DEM) for the solid phase to reasonably simulate the particle behaviors as well as the fluid-particle interactions in multi-phase flows. In this study, in order to validate the multi-fluid model of the SIMMER code coupled with DEM, numerical simulations were performed on a series of self-leveling experiments using a gas injection method in cylindrical particle beds. The effects of friction coefficient on the simulation results were investigated by sensitivity analysis. Though more extensive validations are needed, the reasonable agreement between simulation results and corresponding experimental data preliminarily demonstrates the potential ability of the present method in simulating the self-leveling behaviors of debris bed. It is expected that the SIMMER code coupled with DEM is a prospective computational tool for analysis of safety issues related to solid particle debris bed in SFRs.
Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Uno, Masayoshi*
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10
Heat radiation is one of dominant heat transfer process during a sodium fire event which is a concern in sodium-cooled fast reactor plants. This study aims to model radiation heat transfer from combusting droplets. Radiation energy transport on the combustion flame surface around a sodium droplet is formulated considering emission, absorption and scattering through a similar approach to the formulation of the wall boundary condition. The improved model is tested trough a simple verification analysis and a benchmark analysis on an upward sodium spray combustion experiment. As the result, overestimation of atmospheric temperature and pressure is mitigated by the improved model due to increase in heat transfer to structure.
Bentaib, A.*; Chaumeix, N.*; Grosseuvres, R.*; Bleyer, A.*; Gastaldo, L.*; Maas, L.*; Jallais, S.*; Vyazmina, E.*; Kudriakov, S.*; Studer, E.*; et al.
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
Yamashita, Susumu; Yoshida, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
In order to simulate the relocation phenomena phenomenologically around the reactor core and inside pedestal without assumptions, inputted information, empirical knowledge and given scenarios, a numerical simulation code based on computational fluid dynamics, JUPITER, that can phenomenologically evaluate the melting phenomena has been developed in JAEA. We confirm the applicability of JUPITER to the corium spreading process inside the pedestal by simulating corium spreading behaviors and its distributions under several parameters such as the corium inflow condition. And we investigated detailed fuel debris distribution inside the pedestal and the cavity named sump pit located on the lower part of PCVs, that is distribution for each component of core materials such as stainless steel (SUS), bron carbide (BC), zircaloy (Zry) and uranium dioxide (UO). As a result, since JUPITER uses interface capturing scheme which can treat complicated behavior of interfaces such as a large deformation and a complicated separation among interfaces, the corium was spread with complicated mixing behavior and the distribution of fuel debris tend to accumulate complicatedly inside sump pits. Since existing SA analysis codes difficult to treat such complicated phenomena and the complicated fuel distribution, those result might be contributed to an understanding of circumstances inside PCV and, ultimately, also contributed to reactor decommissioning process.
Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10
The mechanism of Critical Heat Flux (CHF) remains to be clarified, even though it is important to evaluate the CHF for super high heat flux components such as light water reactors (LWRs). Some theoretical models to predict the CHF is proposed so far. A macrolayer formation model which is proposed in order to predict the CHF based on the macrolayer dryout model. In this model, it is assumed that the liquid is captured inside vapor mass at coalescence. In this study, the verification of the assumption of a macrolayer formation model by the numerical simulation of CMFD code, TPFIT, from the view point of hydrodynamics.
Kuwata, Yusuke*; Zhan, Y.*; Enoki, Koji*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10
This study aims improvement of safety analysis for sodium fire accidents in sodium-cooled fast reactors. In the experiment, effect of viscosity on liquid jet impact on solid surface was studied.
Ishigaki, Masahiro; Abe, Satoshi; Shibamoto, Yasuteru; Yonomoto, Taisuke
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
no abstracts in English