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Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako
Nihon Genshiryoku Gakkai-Shi ATOMO
, 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.
Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.666 - 677, 2019/08
An evaluation methodology of critical heat fluxes (CHFs) based on a mechanism for fuel assemblies in light water reactors (LWRs) is needed in order to design and evaluate the safety for the fuel assemblies in LWRs. In our study, the numerical simulation with surface-tracking will be applied for the two-phase flow in fuel assemblies in order to obtain the detail data relating to the size and velocity of bubbles in the subchannel, which is needed to predict the CHF based on the mechanism. In this study, the numerical simulation of two-phase flow in 4
4 bundle was implemented by using JUPITER in order to establish the evaluation method of the size and velocity of bubbles by the numerical simulation, which is the multi-physics simulation code and enable to track the gas-liquid surface. The simulation results are validated by the curve of flow regime for air-water under the adiabatic condition. The bubble and velocity of bubbles obtained by simulation results are analyzed.
Liu, W.; Nariai, Hideki*
Journal of Heat Transfer, 127(2), p.149 - 158, 2005/02
Times Cited Count：12 Percentile：46.02(Thermodynamics)Homogeneous nucleation, although being discounted as a mechanism for vapor formation for water in most conditions, is found being possible to occur under some extreme conditions in subcooled flow boiling. In this paper, firstly, the existence of the homogeneous nucleation governed condition is indicated. Followed, a criterion is developed to judge a given working condition is the conventional one or the homogeneous nucleation governed one. With the criterion, subcooled flow boiling data are categorized and typical homogeneous nucleation governed datasets are listed. CHF triggering mechanism for the homogeneous nucleation governed condition is proposed and verified. Parametric trends of the CHF, in terms of mass flux, pressure, inlet subcooling, channel diameter and the ratio of heated length to diameter are also studied.
Sakai, Mikio; Yamamoto, Toshihiro; Murazaki, Minoru; Miyoshi, Yoshinori
Nuclear Technology, 149(2), p.141 - 149, 2005/02
Times Cited Count：3 Percentile：24.22(Nuclear Science & Technology)In the conventional criticality evaluation of the nuclear powder system, the effects of particulate behavior have not been considered. In other words, it is difficult to reflect the particle behavior into the conventional criticality evaluation. We have developed a novel criticality evaluation code to resolve this issue. The criticality evaluation code, coupling a Discrete Element Method simulation code with a continuous-energy Monte Carlo transport code, makes it possible to study the effect of the particulate behavior on a criticality evaluation. The criticality evaluation code has been applied to the powder system of the MOX fuel powder agitation process. The criticality evaluations have been performed under mixing the MOX fuel powder in a stirred vessel to investigate the effects of the powder boundary deformation and particulate mixture conditions on the criticality evaluation. The evaluation results revealed that the powder uniformity mixture condition and the boundary deformation could make the neutron effective multiplication factor decrease.
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Akimoto, Hajime
Proceedings of Japan-US Seminar on Two-Phase Flow Dynamics, p.317 - 325, 2004/12
We start R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with Power Company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration. In this paper, we will show the R&D plan and describe the current status on experimental and analytical studies. We will confirm the thermal-hydraulic performance in the tight-lattice bundles by this project and develop a predictable technology for the RMWR in future.
Toda, Saburo*; Yuki, Kazuhisa*; Akimoto, Hajime
JAERI-Tech 2004-008, 58 Pages, 2004/03
JAERI-Tech-2004-008.pdf:5.14MB
no abstracts in English
Aso, Tomokazu; Sato, Hiroshi; Kaminaga, Masanori; Hino, Ryutaro; Monde, Masanori*
Proceedings of ICANS-XVI, Volume 2, p.935 - 944, 2003/07
no abstracts in English
Hamada, Kazuya; Koizumi, Norikiyo
Purazuma, Kaku Yugo Gakkai-Shi, 78(7), p.616 - 624, 2002/07
In the Tokamak type fusion reactor design, a forced flow superconducting coil is applied from the viewpoint to high magnetic field, high withstand voltage performance and large electromagnetic force. In the forced flow magnets, it is well known that various electromagnetic phenomena are occurred by zero resistance and diamagnetic effect of superconductor and complicated structure of cable in conduit conductor (CICC). In the R&D of CICC, the study of hysteresis losses and coupling losses CICC have a lot of progress. For example, using the optimization of filament arrangement in superconducting strand and control of contact resistance of strand, ITER model coil project have a large achievement.
Kinoshita, Hidetaka; Nariai, Hideki*; Inasaka, Fujio*
JSME International Journal, Series B, 44(1), p.81 - 89, 2001/01
no abstracts in English
Kato, Takashi
Tabo Kikai, 28(9), p.536 - 545, 2000/09
no abstracts in English
Saito, Yasushi*; Hibiki, Takashi*; Mishima, Kaichiro*; Tobita, Y.*; Suzuki, Toru*; Matsubayashi, Masahito
Proceedings of 9th International Symposium on Flow Visualization, p.391_1 - 391_10, 2000/00
no abstracts in English
Sugimoto, Makoto; Isono, Takaaki; Koizumi, Norikiyo; Nishijima, Gen; Matsui, Kunihiro; Nunoya, Yoshihiko; Takahashi, Yoshikazu; Tsuji, Hiroshi
Cryogenics, 39(11), p.939 - 945, 1999/11
Times Cited Count：3 Percentile：18.87(Thermodynamics)no abstracts in English
Sugimoto, Makoto; Kato, Takashi; Isono, Takaaki; Yoshida, Kiyoshi; Tsuji, Hiroshi
Cryogenics, 39(4), p.323 - 330, 1999/00
Times Cited Count：4 Percentile：23.78(Thermodynamics)no abstracts in English
Sudo, Yukio; Kaminaga, Masanori
Nucl. Eng. Des., 187, p.215 - 227, 1999/00
Times Cited Count：7 Percentile：49.7(Nuclear Science & Technology)no abstracts in English
Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio
Journal of Nuclear Science and Technology, 35(12), p.943 - 951, 1998/12
Times Cited Count：24 Percentile：85.15(Nuclear Science & Technology)no abstracts in English
Sugimoto, Makoto; Isono, Takaaki; Tsuji, Hiroshi; Yoshida, Kiyoshi; *; Hamajima, Takataro*; Sato, Takashi*; Shinoda, K.*
Cryogenics, 38(10), p.989 - 994, 1998/00
Times Cited Count：4 Percentile：23.98(Thermodynamics)no abstracts in English
Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1815 - 1822, 1997/00
In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as "Loss of the primary coolant flow". Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and nonuniform heat flux condition. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed.
; Kukita, Yutaka
Journal of Nuclear Science and Technology, 33(9), p.696 - 702, 1996/09
Times Cited Count：2 Percentile：24.74(Nuclear Science & Technology)no abstracts in English
Onuki, Akira; ; Murao, Yoshio
Journal of Nuclear Science and Technology, 32(3), p.245 - 256, 1995/03
Times Cited Count：1 Percentile：17.53(Nuclear Science & Technology)no abstracts in English
; Kukita, Yutaka; *
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.131 - 136, 1995/00
no abstracts in English
