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Journal Articles

Development of numerical simulation for jet breakup behavior in complicated structure of BWR lower plenum, 6; Influence of the simulant molten fuel properties on jet breakup phenomenon in multi-channels

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.

Journal Articles

Research for thermal-hydraulic performance in a tight-lattice fuel assembly, 2; Two-phase flow analysis in a fuel assembly with a large-scale simulation

Takase, Kazuyuki; Yoshida, Hiroyuki; Akimoto, Hajime; Ose, Yasuo*

Nippon Konsoryu Gakkai Nenkai Koenkai 2005 Koen Rombunshu, p.231 - 232, 2005/08

no abstracts in English

Journal Articles

Numerical visualization on a Large-scale bubbly flow in a vertical small duct

Takase, Kazuyuki; Ose, Yasuo*; Yoshida, Hiroyuki; Akimoto, Hajime; Aoki, Takayuki*

Kashika Joho Gakkai-Shi, 25(Suppl.1), p.435 - 436, 2005/07

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large-scale numerical simulation, 4; Large-scale analysis of water-vapor two-phase flow in rod bundles with TPFIT code using earth simulator

Yoshida, Hiroyuki; Ose, Yasuo*; Kureta, Masatoshi*; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.106 - 114, 2005/06

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large-scale numerical simulation, 3; Analysis of liquid film falling down on inclined flat plate

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(1), p.25 - 31, 2005/03

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large-scale numerical simulation, 2; Experimental analysis of 2-channel fluid mixing tests

Nagayoshi, Takuji*; Yoshida, Hiroyuki; Onuki, Akira; Akimoto, Hajime

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(1), p.16 - 24, 2005/03

A detailed gas-liquid two-phase flow analysis code based on an advanced interface-tracking method has been developed. It is expected that the developed code would be able to simulate two-phase cross flow behavior within tight-lattice fuel bundles without relying on any empirical correlations. In order to verify the applicability of the code to simulate two-phase cross flow behavior in such situations, numerical analyses of 2-channel model tests were conducted to compare the air slug deformation and separation behavior caused by cross flow through a narrow interconnection between channels. Although the code underestimated the ascending velocity of the slug, the calculated slug deformation and separation behavior were shown to be quite similar to those observed by a high-speed video camera. Moreover the minimum differential pressure between the subchannels through the interconnection, causing channel-to-channel air transfer to occur could be predicted to within 20Pa. However, further studies of modeling and implementation related to the interface-channel wall interaction, such as a contact angle of a gas-liquid interface at the channel wall, are required for prediction improvements. Nevertheless, the qualitative capability of the developed code to simulate two-phase cross flow phenomena was demonstrated.

Journal Articles

Numerical simulation of single bubble behavior in rod bundle with interface tracking method

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Tamai, Hidesada; Takase, Kazuyuki; Akimoto, Hajime

Proceedings of 4th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-4), p.264 - 269, 2004/12

no abstracts in English

Oral presentation

Oral presentation

Numerical simulation of gas entrainment phenomena with interface-tracking method

Ito, Kei; Koizumi, Yasuo*; Ohshima, Hiroyuki; Kawamura, Takumi*

no journal, , 

The authors are developing a high-precision CFD code with an enhanced interface tracking method to simulate the gas entrainment (GE) phenomena in sodium-cooled fast reactors (SFRs). The GE in SFRs is characterized by an elongated interfacial dent (gas core) along the vortex core and the bubble pinch-off at the tip of the gas core. The authors have developed physics-basis algorithms which model accurately the interfacial shape (gradient and curvature) and the mechanical balance condition at an interface. In this paper, a simple experiment of the GE is simulated to validate the developed code. The simulation result of the bubble pinch-off and the entrained flow rate shows good agreement with the experimental data, that is, the developed simulation code can be applicable to the evaluation of the GE in SFRs.

Oral presentation

Numerical simulation of two-phase flow in 4$$times$$4 simulated fuel bundle using TPFIT

Ono, Ayako; Nagatake, Taku; Suzuki, Takayuki*; Yoshida, Hiroyuki

no journal, , 

The evaluation method for the critical heat flux based on a mechanism is needed to evaluate the safety of fuel bundles in a light water reactor. The development of the numerical simulation method to predict the two-phase flow in the fuel bundles is implemented, which is needed to predict the CHF. The bubbly flow in 4$$times$$4 simulated fuel bundle was calculated in order to develop the evaluation method of the two-phase flow in the fuel bundle using TPFIT. The applicability of TPFIT for the two-phase flow in the fuel bundle was confirmed.

Oral presentation

Large-scale simulation on two-phase flow in the fuel assemblies of LWR by the mechanistically based method

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

no journal, , 

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. The simulation results and previous study relating to the observation of two-phase flow in fuel bundle are compared to validate the simulation result and extract the issue to be considered.

Oral presentation

Numerical simulation on two-phase flow in the 4$$times$$4 fuel bundle with the spacer by the mechanistically based method

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

no journal, , 

JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code in order 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. In this study, the JUPITER code is examined to apply for the two-phase flow simulation of 4$$times$$4 fuel assembly with the spacer grid. The simulation results and previous study relating to observation of the two-phase flow in the fuel bundle are compared to validate the simulation results and considered reproducibility of the effect of the spacer grid.

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