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Journal Articles

Numerical simulation of two-phase flow in fuel assemblies with a spacer grid using a mechanistically based method

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.

Journal Articles

Toward mechanistic evaluation of critical heat flux in nuclear reactors, 2; Recent studies and future challenges toward mechanistic and reliable CHF evaluation

Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(12), p.820 - 824, 2021/12

The evaluation method of the critical heat flux based on the mechanism is needed for the efficient design and development of fuel in reactors and the appropriate safety evaluation. In this paper, the current researches relating to the mechanism of the critical heat flux are reviewed, and the issue to be considered in the future are discussed.

Journal Articles

Measurement of void fraction distribution in steam-water two-phase flow in a 4$$times$$4 bundle at 2 MPa

Liu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki

Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06

To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 $$times$$ 4 bundle under 1.6 MPa (202 $$^{circ}$$C), 2.1 MPa (215 $$^{circ}$$C) and 2.6 MPa (226 $$^{circ}$$C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.

Journal Articles

Current status of thermal/hydraulic feasibility project for reduced-moderation water reactor, 2; Development of two-phase flow simulation code with advanced interface tracking method

Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04

The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

Journal Articles

A Large-scale numerical simulation of bubbly and liquid film flows in narrow fuel channels

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

Proceedings of 2005 ASME International Mechanical Engineering Congress and Exposition (CD-ROM), 8 Pages, 2005/11

no abstracts in English

Journal Articles

Predicted two-phase flow structure in a fuel bundle of an advanced light-water reactor

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10

no abstracts in English

Journal Articles

Investigation of water-vapor two-phase flow characteristics in a tight-lattice core by large scale numerical simulation, 1; Development of a direct analysis procedure on two-phase flow with an advanced interface tracking method

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

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09

When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.

Journal Articles

Numerical simulation on large-scale bubbly flow behavior in a narrow duct

Takase, Kazuyuki; Yoshida, Hiroyuki; Tamai, Hidesada; Ose, Yasuo*

Nihon Kikai Gakkai 2004-Nendo Nenji Taikai Koen Rombunshu, Vol.2 (No.04-1), p.251 - 252, 2004/09

no abstracts in English

Journal Articles

Numerical analysis of a water-vapor two-phase film flow in a narrow coolant channel with a three-dimensional rectangular rib

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada

JSME International Journal, Series B, 47(2), p.323 - 331, 2004/05

no abstracts in English

Journal Articles

Three-dimensional computations of two-phase flow behavior in a simulated fusion reactor under water ingress

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

Proceedings of the 1st International Symposium on Advanced Fluid Information (AFI-2001), p.227 - 232, 2001/10

no abstracts in English

Journal Articles

Three-dimensional void fraction measurement of two-phase flow in a rod bundle by neutron radiography

*; *; Fujii, Terushige*; *; Matsubayashi, Masahito; Tsuruno, Akira

Fifth World Conf. on Neutron Radiography, 0, p.118 - 125, 1996/00

no abstracts in English

Journal Articles

Interfacial friction factor for high-pressure steam/water stratified-wavy flow in horizontal pipe

Nakamura, Hideo; Kukita, Yutaka; *

Journal of Nuclear Science and Technology, 32(9), p.868 - 879, 1995/09

 Times Cited Count:4 Percentile:43.29(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Flow regime transition to wavy dispersed flow for high-pressure steam/water two-phase flow in horizontal pipe

Nakamura, Hideo; Kukita, Yutaka; *

Journal of Nuclear Science and Technology, 32(7), p.641 - 652, 1995/07

 Times Cited Count:5 Percentile:49.7(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Investigation of centrifugal pump performance under two-phase flow conditions

G.R.Noghrehkar*; Kawaji, Masahiro*; A.M.C.Chan*; Nakamura, Hideo; Kukita, Yutaka

J. Fluids Eng., 117, p.1 - 9, 1995/03

no abstracts in English

Journal Articles

Visualization of fluid phenomena using a high frame-rate neutron radiography with a steady thermal neutron beam

Hibiki, Takashi*; Mishima, Kaichiro*; Yoneda, Kenji*; Fujine, Shigenori*; Tsuruno, Akira; Matsubayashi, Masahito

Nuclear Instruments and Methods in Physics Research A, 351, p.423 - 436, 1994/00

 Times Cited Count:36 Percentile:92.88(Instruments & Instrumentation)

no abstracts in English

Journal Articles

Fiber-optics video probes for observation of high-pressure high-temperature two-phase flow

Nakamura, Hideo; Murata, Hideo; Ito, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Kukita, Yutaka

Kashika Joho Gakkai-Shi, 12(47), p.47 - 56, 1992/10

no abstracts in English

Journal Articles

Fiber-optics video probes for observation of high-pressure two-phase flow

Nakamura, Hideo; Murata, Hideo; Ito, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Kukita, Yutaka

ANS Proc. 1991 National Heat Transfer Conf., Vol. 5, p.175 - 180, 1991/00

no abstracts in English

Journal Articles

Hydrodynamics of ECC water bypass and refill of lower plenum at PWR-LOCA

*; Murao, Yoshio

Journal of Nuclear Science and Technology, 24(10), p.785 - 797, 1987/10

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

no abstracts in English

29 (Records 1-20 displayed on this page)