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Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
The new prediction method of critical heat flux (CHF) of the fuel assemblies based on the mechanism is proposed in this study. The prediction method of CHF based on the mechanism has been needed for a long time to enhance the safety analysis and reduce the design cost. From several experimental findings of the liquid-vapor behavior near the heating surface from the nucleate boiling to the CHF, the authors consider that the macrolayer dryout model will be appropriate to predict the CHF under the reactor condition. The prediction method of the macrolayer thickness and the passage period of vapor mass in the fuel assemblies are needed to predict CHF from the macrolayer dryout model. In this study, the CHF under the forced convection is evaluated by combining the prediction methods for the macrolayer thickness and passage period of vapor mass, which are proposed by authors. The prediction of the CHF under the forced convection is examined and compared with the experimental data.
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.
Ono, Ayako; Yamashita, Susumu; Sakashita, Hiroto*; Suzuki, Takayuki*; Yoshida, Hiroyuki
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
JAEA is implementing a simulation of a two-phase flow in the reactor core by TPFIT and JUPITER which are developed by JAEA based on the surface tracking method. However, it is impossible to simulate a boiling on the heating surface in the large-scale domain by this type of simulation method since the simulation of boiling based on the surface tracking method needs the fine mesh which sufficiently resolves the initiation of boiling. Therefore, JAEA started to develop the simplified boiling model applied for the two-phase flow in the fuel assemblies. In this study, the simulation results of the convection boiling on a vertical heating surface and the comparison between the simulation results and experimental results are shown.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.
Suzuki, Takayuki*; Yoshida, Hiroyuki; Horiguchi, Naoki; Yamamura, Sota*; Abe, Yutaka*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
4 simulated bundleOno, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06
JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08
An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. In our study, the numerical simulation with surface-tracking will be applied for the two-phase flow in fuel assemblies in order to obtain the detail data relating to the size and velocity of bubbles in the subchannel, which is needed to predict the CHF based on the mechanism. In this study, the numerical simulation of two-phase flow in 44 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. The simulation results are validated by the curve of flow regime for air-water under the adiabatic condition. The bubble and velocity of bubbles obtained by simulation results are analyzed.
Uesawa, Shinichiro; Horiguchi, Naoki; Suzuki, Takayuki*; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
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.
Uesawa, Shinichiro; Suzuki, Takayuki*; Yoshida, Hiroyuki
Konsoryu Shimpojiumu 2018 Koen Rombunshu (Internet), 2 Pages, 2018/08
no abstracts in English
Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
The mechanism of critical heat flux (CHF) for higher system pressure remains to be clarified, even though it is important to evaluate the CHF for the light water reactor (LWR) which is operated under the high pressure condition. In this study, the process of bubble coalescence was simulated by using a computational multi-fluid dynamics (CMFD) simulation code TPFIT under various system pressure in order to investigate the behavior of bubbles as a basic study. The growth of bubbles was simulated by blowing of vapor from a tiny orifice simulating bubble bottom. One or four orifices were located on the bottom surface in this simulation study. The numerical simulations were conducted by varying the pressure and temperature.
Suzuki, Takayuki; Yoshida, Hiroyuki; Abe, Yutaka*; Kaneko, Akiko*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred. Especially, it is important to estimate the behavior of molten core jet in the lower part of the reactor pressure vessel at a severe accident. In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes, instrument guide tubes and core support plate. The objective of this study is to develop the simulation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Therefore, experimental works by use of multi-phase flow visualization technique were also carried out. In the experiments, time series of interface shapes are observed by high speed camera and velocity profiles in/out of the jet were measured by the PIV method. In this paper, we carried out a numerical simulation of the jet breakup phenomena in the multi-channels with various simulant molten materials to evaluate the influence of properties on the jet breakup phenomena. As a result, it was confirmed that density and surface tension affected on the falling down velocity of the simulant materials and the interface behavior of the molten jet. However, viscosities of the simulant materials have small effects on jet breakup phenomena, including the interface shape and size of fragments.
Suzuki, Takayuki; Yoshida, Hiroyuki; Nagase, Fumihisa
Journal of Nuclear Science and Technology, 51(7-8), p.968 - 976, 2014/07
Times Cited Count：7 Percentile：48.36(Nuclear Science & Technology)Saito, Ryusuke*; Abe, Yutaka*; Kaneko, Akiko*; Suzuki, Takayuki; Yoshida, Hiroyuki; Nagase, Fumihisa
Nihon Kikai Gakkai Kanto Shibu Dai-19-Ki Sokai Koenkai Koen Rombunshu, p.421 - 422, 2013/03
It is important to estimate the behavior of molten core jet in RPV (Reactor Pressure Vessel) of BWR when severe accident occurred as can be seen from Fukushima Daiichi Nuclear Power Plant accident. Thus we are developing a simulation code to estimate the behavior of molten core falling in a lower plenum of BWR. Since there are many complicated structures in the plenum, it is expected that molten core would be injected in the complicated flow channel. The objective of the present study is to investigate the influence of complicated structures in a lower plenum on jet by experiment. To investigate it, we observe jet injection which simulated the molten core falling accident experimentally. We constructed a model of a lower plenum and a steady jet injection equipment. Results of experiment showed that jet behavior is affected by complicated flow structure.
Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki; Ikuta, Ryuhei*; Koizumi, Yasuo*
Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.103 - 106, 2011/06
no abstracts in English
Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11
Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki; Koizumi, Yasuo*
Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 4 Pages, 2010/10
Yoshida, Hiroyuki; Suzuki, Takayuki*; Takase, Kazuyuki
Nihon Konsoryu Gakkai Nenkai Koenkai 2010 Koen Rombunshu, p.344 - 345, 2010/07
no abstracts in English
Yoshida, Hiroyuki; Hosoi, Hideaki; Suzuki, Takayuki*; Takase, Kazuyuki
Dai-15-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.273 - 276, 2010/06
no abstracts in English
Yoshida, Hiroyuki; Hosoi, Hideaki*; Suzuki, Takayuki*; Takase, Kazuyuki
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
