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Journal Articles

CFD analysis

Takase, Kazuyuki; Misawa, Takeharu*

Supercritical-Pressure Light Water Cooled Reactors, p.301 - 319, 2014/12

no abstracts in English

Journal Articles

Numerical prediction on heat transfer-characteristics of supercritical pressure water in a heated pipe based on three dimensional two-fluid model

Ose, Yasuo*; Yoshimori, Hajime*; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Nihon Kikai Gakkai Dai-26-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.701_1 - 701_2, 2013/11

no abstracts in English

Journal Articles

Numerical evaluation of influence of coolant flow direction upon thermal hydraulics in SCWR rod bundle

Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Nihon Kikai Gakkai Dai-26-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.702_1 - 702_2, 2013/11

no abstracts in English

Journal Articles

A Large-scale three-dimensional simulation on thermal-hydraulics in a fuel bundle for SCWR

Misawa, Takeharu; Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Oka, Yoshiaki*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo (SNA & MC 2013) (CD-ROM), 2 Pages, 2013/10

Journal Articles

Numerical analysis of heat transfer enhancement by spacers in subchannels at supercritical pressure condition

Takase, Kazuyuki; Misawa, Takeharu; Yoshida, Hiroyuki

Proceedings of 4th International Conference on Jets, Wakes and Separated Flows (ICJWSF 2013) (CD-ROM), 6 Pages, 2013/09

For the thermal design of supercritical water reactors, it is necessary to develop an analysis method to correctly predict turbulent heat transfer characteristics in subchannels of fuel bundles. Spacers are set into the subchannels to keep a distance between adjacent fuel rods. The turbulent heat transfer generally enhances by reduction of cross-sectional area in the subchannels due to the existence of the spacers. However, since thermo-physical properties of supercritical fluids drastically change at the vicinity of the pseudocritical point, the enhancement of the turbulent heat transfer depends on the thermal design. Then, the Japan Atomic Energy Agency is developing the analysis method to predict thermal-hydraulic characteristics of the supercritical fluids. The heat transfer calculations were performed using a newly developed code under the condition of a subchannel with a specer, and the enhancement of the turbulent heat transfer coefficient in subchannels with the spacer was analyzed numerically.

Journal Articles

Status of experimental and analytical thermal-hydraulic studies on severe accident events at Fukushima Dai-ichi NPP

Takase, Kazuyuki; Yoshida, Hiroyuki; Liu, W.; Misawa, Takeharu; Nagatake, Taku; Yamashita, Susumu

Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 6 Pages, 2013/04

Journal Articles

Numerical predictions of crossing-flow characteristics of a supercritical fluid in parallel circular ducts with a narrow rectangular channel at isothermal condition

Takase, Kazuyuki; Misawa, Takeharu; Yoshida, Hiroyuki; Mori, Hideo*

Proceedings of 6th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-6) (CD-ROM), 9 Pages, 2013/03

Numerical analyses of crossing flows between two parallel circular channels were conducted for a specific geometry that simply modeled subchannels in a fuel bundle of a supercritical water reactor. The two parallel circular channels were connected by a rectangular channel in the axial direction. Crossing flow occurred in the rectangular channel and was caused by differences in temperatures of fluids flowing in the two channels. The working fluid was supercritical Freon. The SST turbulence model was chosen for precisely calculating the boundary layers of temperature and velocity near the channel walls. From the analytical results, relations between crossing flow and fluid temperature were clarified quantitatively.

Journal Articles

Numerical predictions on turbulent mixing in two parallel channels with a connecting passage at non-heated supercritical fluid condition

Misawa, Takeharu; Takase, Kazuyuki; Mori, Hideo*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 6 Pages, 2012/12

Journal Articles

Crossing flow simulation in two parallel channels by supercritical fluids

Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Nihon Kikai Gakkai Dai-25-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.726 - 727, 2012/10

no abstracts in English

Journal Articles

Numerical analysis on a spacer effect in subchannels of fuel bundles of supercritical water reactors

Takase, Kazuyuki; Misawa, Takeharu; Yoshida, Hiroyuki

Nihon Kikai Gakkai Dai-25-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.723 - 725, 2012/10

no abstracts in English

Journal Articles

Journal Articles

Numerical research on turbulent mixing and crossing flow characteristics under supercritical fluid conditions

Takase, Kazuyuki; Misawa, Takeharu; Yoshida, Hiroyuki

Proceedings of Advances in Thermal Hydraulics (ATH '12) (CD-ROM), p.23 - 28, 2012/01

no abstracts in English

Journal Articles

Thermal-hydraulic calculation for simplified fuel assembly of super fast reactor using two-fluid model analysis code ACE-3D

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 5 Pages, 2012/00

In the present paper, thermal-hydraulic behavior in a simplified fuel assembly of the supercritical water cooled fast reactor (Super Fast Reactor) was analyzed with the three-dimensional two-fluid model analysis code ACE-3D. The analytical geometry simulates a 19-rod assembly, which is one of the most simplified geometry of the SCWR fuel assembly and includes three kinds of different subchannel types; (1) adjoining to the channel box, (2): next to type (1), and (3): located inside types (1) and (2). It was confirmed that the MCST satisfies a thermal design criteria to ensure fuel and cladding integrity.

Journal Articles

Super fast reactor R&D projects in Japan, 4; Numerical estimation of thermal-hydraulic characteristics of supercritical fluids in tight-lattice bundles by three-dimensional two-fluid model analysis code ACE-3D

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 5 Pages, 2011/10

To analyze thermal hydraulics in the core of supercritical-water-cooled reactors, JAEA has been improved a three-dimensional two-fluid model analysis code ACE-3D, which has been developed originally for two-phase flow of LWRs. Heat transfer experiments of supercritical fluids flowing in a tube, a vertical annular channel around a heater pin and 7-rod bundles were analyzed with the improved ACE-3D to assess the prediction performance of the code at supercritical region. As a result, it was confirmed that the calculated wall surface temperatures agreed with the measured results. To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of Super Fast Reactor, a simplified 19-rod fuel assembly was analyzed. Maximum clad surface temperature was observed at the position facing to the narrowest gap on the center rod near the outlet and the value was 901K. The predicted MCST satisfies thermal design criteria to ensure fuel and cladding integrity.

Journal Articles

Numerical analysis on thermal-hydraulics of supercritical water flowing in a tight-lattice fuel bundle

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Progress in Nuclear Science and Technology (Internet), 2, p.143 - 146, 2011/10

To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of supercritical water reactor (Super Fast Reactor), a simplified 19-rod fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D. In this calculation, a one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power distribution are to be the same as the steady state condition of the reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Effect of grid spacers is taken into account in the analysis. Maximum cladding surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 628$$^{circ}$$C that is almost the same as results without grid spacers. The predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650$$^{circ}$$C.

Journal Articles

A Numerical prediction on pressure drop characteristics in a fuel bundle of an accelerator driven system

Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Dai-16-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.109 - 110, 2011/06

no abstracts in English

Journal Articles

Assessment of applicability of two-fluid model code ACE-3D to heat transfer test of supercritical water flowing in an annular channel

Nakatsuka, Toru; Ezato, Koichiro; Misawa, Takeharu; Seki, Yohji; Yoshida, Hiroyuki; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Takase, Kazuyuki

Journal of Nuclear Science and Technology, 47(12), p.1118 - 1123, 2010/12

 Times Cited Count:1 Percentile:11(Nuclear Science & Technology)

In order to perform efficiently the thermal design of the supercritical water reactor (SCWR), it is important to assess the thermal hydraulics in rod bundles of the core. Japan Atomic Energy Agency (JAEA) has been improved the three-dimensional two-fluid model analysis code ACE-3D, which has been developed originally for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water at supercritical region. In the present paper, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which was performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code may be applicable to prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of SCWR core.

Journal Articles

Numerical visualization of boiling two-phase flow behavior in fuel bundles at simulated earthquake condition

Takase, Kazuyuki; Misawa, Takeharu; Yoshida, Hiroyuki

Kashika Joho Gakkai-Shi, 30(Suppl.2), p.25 - 26, 2010/10

In order to evaluate an influence of earthquake acceleration to the boiling two-phase flow behavior in nuclear reactors, numerical simulations were performed under the simulated earthquake condition. The two-phase flow analysis code, ACE-3D, was modified as the influence of the earth quake acceleration can calculate. To check out if the modification is adequate, a series of calculations were carried out and the following summaries were derived; (1) the void fraction in the fuel bundle receives the influence of the earthquake, (2) the liquid-phase in the two-phase flow moves in the same direction as the direction of oscillation due to the inputted earthquake acceleration, and (3) due to the density difference in comparison with the liquid phase, the gas phase of that moves in the direction opposite to the oscillating direction. This study enabled visualized evaluation of the boiling two-phase flow behavior in the nuclear reactors at the earthquake condition.

Journal Articles

Numerical simulation of boiling two-phase flow in a simulated subchannel of fuel assemblies excited by earthquake oscillation

Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 6 Pages, 2010/10

Journal Articles

Numerical analysis on thermal-hydraulics of supercritical water flowing in a tight-lattice fuel bundle

Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

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

To evaluate thermal hydraulic characteristics of a tight-lattice fuel bundle of supercritical water reactor (Super Fast Reactor), a simplified 19-rod fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D. In this calculation, a one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power distribution are to be the same as the steady state condition of the reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Effect of grid spacers is taken into account in the analysis. Maximum clad surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 628$$^{circ}$$C that is almost the same as results without grid spacers. The predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650$$^{circ}$$C.

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