CFD analysis

Takase, Kazuyuki; Misawa, Takeharu*

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

https://doi.org/10.1007/978-4-431-55025-9_3

no abstracts in English

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

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.

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.

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

Development of an analytical method on water-vapor boiling two-phase flow characteristics in BWR fuel assemblies under earthquake condition

Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki

Nuclear Reactors, p.157 - 174, 2012/02

https://doi.org/10.5772/32299

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

http://www-pub.iaea.org/mtcd/meetings/PDFplus/2009/cn176/cn176_BoeS.pdf

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.

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

https://doi.org/10.15669/pnst.2.143

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 628C 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 650C.

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

https://doi.org/10.3327/jnst.47.1118

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.

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 628C 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 650C.