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, 63(12), p.820 - 824, 2021/12

https://doi.org/10.3327/jaesjb.63.12_820

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.

Development of numerical simulation method for capturing behavior of aerosol particles on gas-liquid interface based on interface tracking method

Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*

Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06

no abstracts in English

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

In-beam fissio study at JAEA for heavy element synthesis

Nishio, Katsuhisa; Ikezoe, Hiroshi; Hofmann, S.*; Ackermann, D.*; Aritomo, Yoshihiro*; Comas, V. F.*; Dllmann, Ch. E.*; Heinz, S.*; Heredia, J. A.*; Heberger, F. P.*; et al.

AIP Conference Proceedings 1524, p.68 - 72, 2013/04

https://doi.org/10.1063/1.4801679

Fusion probabilities in the reactions Ca + U at energies around the Coulomb barrier

Nishio, Katsuhisa; Mitsuoka, Shinichi; Nishinaka, Ichiro; Makii, Hiroyuki; Wakabayashi, Yasuo*; Ikezoe, Hiroshi; Hirose, Kentaro*; Otsuki, Tsutomu*; Aritomo, Yoshihiro*; Hofmann, S.*

Physical Review C, 86(3), p.034608_1 - 034608_6, 2012/09

https://doi.org/10.1103/PhysRevC.86.034608

Nematic-to-smectic transition of magnetic texture in conical state

Takeda, Masayasu; Endo, Yasuo; Kakurai, Kazuhisa; Onose, Yoshinori*; Suzuki, Junichi; Tokura, Yoshinori*

Journal of the Physical Society of Japan, 78(9), p.093704_1 - 093704_4, 2009/09

https://doi.org/10.1143/JPSJ.78.093704

The complex phase transition of the magnetic long-range order in FeCoSi has been investigated by small-angle neutron scattering (SANS) measurement at low temperatures under an external magnetic field. A variety of scattering patterns, such as ring, crescent-shaped, and diffusive spots, were observed by changing temperature and applying a magnetic field. The results could be interpreted by an analogy of the phase transition of the liquid crystal when the net magnetization is regarded as the director of the liquid crystal.

Large-scale simulations on thermal-hydraulics in fuel bundles of advanced nuclear reactors

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Kano, Takuma; Merzari, E.*; Ninokata, Hisashi*

Annual Report of the Earth Simulator Center April 2006 - March 2007, p.223 - 228, 2007/09

http://www.jamstec.go.jp/esc/publication/annual/annual2006/

no abstracts in English

Study on cross flow phenomena in a tight-lattice rod bundle by statistical method

Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*

Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.85 - 88, 2007/06

As a candidate for next generation reactor, the innovative FLexible-fuel-cycle Water Reactor (FLWR) adopts a remarkably tight triangular lattice arrangement with about 1 mm gap spacing between adjacent fuel rods. In relation to its design, this study presents a statistical evaluation of numerical simulation results of a detailed two-phase flow simulation code (named TPFIT). In order to make clear mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is used to simulate cross flow between two modeled subchannels. Attention was focused on instantaneous fluctuation characteristics of differential pressure between two subchannels and gas/liquid mixing coefficients. With the calculation of correlation coefficients between the differential pressure and gas/liquid mixing coefficients, the time scales of cross flow, e.g. lag times were evaluated, and the effects of mixing section length, flow pattern and gap spacing on correlation coefficients were extensively investigated. The difference in mechanism between gas and liquid cross flows was pointed out.

A Large-scale simulation on water-vapor bubbly flow dynamics in fuel bundles of advanced nuclear reactors

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

Annual Report of the Earth Simulator Center April 2005 - March 2006, p.261 - 265, 2007/01

http://www.jamstec.go.jp/esc/publication/annual/annual2005/

no abstracts in English

A Large-scale simulation on two-phase flow characteristics around duel rods in a tight-lattice core

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

Proceedings of 2006 ASME International Mechanical Engineering Congress & Exposition (IMECE 2006) (CD-ROM), 8 Pages, 2006/11

Water-vapor two-phase flow structure in a fuel bundle of an advanced light water reactor was analyzed numerically by large-scale direct simulations. A newly developed two-phase flow analysis code was used. It can precisely predict the interface behavior between the liquid and gas phase by using the interface tracking method. The present analytical geometry simulates a tight-lattice fuel bundle with 37 fuel rods and four spacers. The fuel rod outer diameter is 13 mm and gap spacing between each rod is 1.3 mm. Each spacer is installed in an arbitrary axial position in order to keeping the gap width. Water flows upward from the bottom of the fuel bundle. The inlet conditions of water are as follows: temperature 283C, pressure 7.2 MPa, flow rate 400 kg/ms, and the Reynolds number 40,000. In the present study three-dimensional computations were carried out under the non-heated isothermal flow condition in order to remove the effect of heat transfer by the fuel rods. The average mesh size in the present numerical study was 0.15 mm. From results of a series of the numerical simulations, the following consideration was derived: (1) The fuel rod surface is encircled with thin water film; (2) The bridge phenomenon by the water film appears in the region where the spacing between fuel rods is narrow; (3) Vapor flows downward the triangular region where the spacing between fuel rods is large; and, (4) A flow configuration of vapor shows the streak structure in the vertical direction.