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Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*; Tsukamoto, Naofumi*; Kaneko, Junichi*
JAEA-Review 2024-039, 45 Pages, 2024/09
JAEA-Review-2024-039.pdf:2.23MB
Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Therefore, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing-flow mixed convection. Furthermore, the authors evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations can accurately predict all the experimental data. The authors confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids.
funabiki, Yuta*; Iyota, Muneyoshi*; Shobu, Takahisa; Matsuda, Tomoki*; Hayashi, Yujiro*; Sano, Tomokazu*; 8 of others*
Journal of Manufacturing Processes, 115, p.40 - 55, 2024/04
Times Cited Count：3 Percentile：72.20(Engineering, Manufacturing)Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
For contaminated water management in decommissioning Fukushima Daiichi Nuclear Power Stations, reduction in water injection, intermittent injection water and air cooling are considered. However, since there are uncertainties of fuel debris in the PCV, it is necessary to examine and evaluate optimal cooling methods according to the distribution state of the fuel debris and the progress of the fuel debris retrieval work in advance. We have developed a method for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation and the porosity of fuel debris. Since a large-scale thermal-hydraulics analysis of natural convection is necessary for the method, JUPITER developed independently by JAEA is used. It is however difficult to perform the large-scale thermal-hydraulics analysis with JUPITER by modeling the internal structure of the debris which may consist of a porous medium. Therefore, it is possible to analyze the heat transfer of the porous medium by adding porous models to JUPITER. In this study, we report the validation of JUPITER applied the porous model and discuss which heat transfer models are most effective in porous models such as series, parallel and geometric mean models. To obtain validation data of JUPITER for the natural convective heat transfer analysis around the porous medium, we performed the heat transfer and the flow visualization experiments of the natural convection in the experimental system including the porous medium. In the comparison between the experiment and the numerical analysis with each model, the numerical result with the geometric mean model was the closest of the models to the experimental results. However, the numerical results of the temperature and the velocity were overestimated for those experimental results. In particular, the temperature near the interface between the porous medium and air was more overestimated.
Ono, Ayako; Sakashita, Hiroto*; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
The new prediction method of critical heat flux (CHF) of the fuel assemblies based on the mechanism is proposed in this study. The prediction method of CHF based on the mechanism has been needed for a long time to enhance the safety analysis and reduce the design cost. From several experimental findings of the liquid-vapor behavior near the heating surface from the nucleate boiling to the CHF, the authors consider that the macrolayer dryout model will be appropriate to predict the CHF under the reactor condition. The prediction method of the macrolayer thickness and the passage period of vapor mass in the fuel assemblies are needed to predict CHF from the macrolayer dryout model. In this study, the CHF under the forced convection is evaluated by combining the prediction methods for the macrolayer thickness and passage period of vapor mass, which are proposed by authors. The prediction of the CHF under the forced convection is examined and compared with the experimental data.
Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
nard convection using new developed mist modelsOhira, Hiroaki*; Tanaka, Masaaki; Yoshikawa, Ryuji; Ezure, Toshiki
Annals of Nuclear Energy, 172, p.109075_1 - 109075_10, 2022/07
Times Cited Count：1 Percentile：16.36(Nuclear Science & Technology)In order to evaluate the mist behavior in the cover gas region of Sodium-cooled Fast Reactors (SFRs) in good accuracy, turbulent model for Rayleigh-Bnard convection (RBC) was selected, and the Reynolds-averaged number density and momentum equations for mist behavior were developed and incorporated into the OpenFOAM code. In the first stage, the RBC in a simple parallel channel was calculated using Favre-averaged k-
SST model. The average temperature and flow characteristics agreed well with results from DNS, LES, and experiments. Then the basic heat transfer experiment simulating the cover gas region of SFRs was calculated using this turbulent model and new mist models. The calculated average temperature distribution in the height direction and the mist mass concentration agreed well with the experimental results. We developed a method that could simulate the mist behavior in turbulent RBC environments and the cover gas region of SFRs with high accuracy.
Ishigaki, Masahiro*; Abe, Satoshi; Hamdani, A.; Hirose, Yoshiyasu
Annals of Nuclear Energy, 168, p.108867_1 - 108867_20, 2022/04
Times Cited Count：5 Percentile：64.62(Nuclear Science & Technology)Abe, Satoshi; Hamdani, A.; Ishigaki, Masahiro; Shibamoto, Yasuteru
Proceedings of International Topical Meeting on Advances in Thermal Hydraulics (ATH 2020) (Internet), p.258 - 268, 2020/10
Wang, Z.; Duan, G.*; Matsunaga, Takuya*; Sugiyama, Tomoyuki
International Journal of Heat and Mass Transfer, 157, p.119919_1 - 119919_20, 2020/08
Times Cited Count：23 Percentile：79.31(Thermodynamics)Takada, Shoji; Ngarayana, I. W.*; Nakatsuru, Yukihiro*; Terada, Atsuhiko; Murakami, Kenta*; Sawa, Kazuhiro*
Mechanical Engineering Journal (Internet), 7(3), p.19-00536_1 - 19-00536_12, 2020/06
In this study reasonable 2D model was established by using FLUENT for start-up of analysis and evaluation of heat transfer flow characteristics in 1/6 scale model of VCS for HTTR. By setting up pressure vessel temperature around 200
C about relatively high ratio of heat transfer via natural convection in total heat removal around 20-30%, which is useful for code to experiment benchmark in the aspect to confirm accuracy to predict temperature distribution of components which is heated up by natural convection flow. The numerical results of upper head of pressure vessel by the 
--SST intermittency transition model, which can adequately reproduce the separation, re-adhesion and transition, reproduced the test results including temperature distribution well in contrast to those by the -
model in both cases that helium gas is evacuated or filled in the pressure vessel. It was emerged that any local hot spot did not appear on the top of upper head of pressure vessel where natural convection flow of air is separated in both cases. In addition, the plume of high temperature helium gas generated by the heating of heater was well mixed in the upper head and uniformly heated the inner surface of upper head without generating hot spots.
Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.
JAEA-Technology 2019-008, 12 Pages, 2019/07
JAEA-Technology-2019-008.pdf:2.37MB
As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.
Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Morita, Koji*
Annals of Nuclear Energy, 96, p.137 - 147, 2016/10
Times Cited Count：5 Percentile：40.14(Nuclear Science & Technology)After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Specifically, emergency power generators are not necessary and the decay heat can be passively removed for a long time, even forever if the heat removal capacity of the RCCS is sufficient. We can also define the experimental conditions on radiation and natural convection for the scale-down heat removal test facility.
Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki
Heat Transfer-Asian Research, 34(5), p.293 - 308, 2005/07
Water cooling panels have been adopted as the vessel cooling system of the HTTR to cool the reactor core indirectly by natural convection and thermal radiation. In order to investigate the heat transfer characteristics of high temperature gas in a vertical annular space between the reactor pressure vessel and cooling panels of the HTTR, we carried out experiments and numerical analyses on natural convection heat transfer coupled with thermal radiation heat transfer in an annulus between two vertical concentric cylinders with the inner cylinder heated and the outer cylinder cooled. In the present experiments, Rayleigh number based on the height of the annulus ranged from 2.0
10
to 5.410 for helium gas and from 1.210
to 3.510 for nitrogen gas. The numerical results were in good agreement with the experimental ones regarding the surface temperatures of the heating and cooling walls. As a result of the experiments and the numerical analyses, the heat transfer coefficient of natural convection coupled with thermal radiation was obtained.
Fumizawa, Motoo; Tanaka, Gaku*; Zhao, H.*; Hishida, Makoto*; Shiina, Yasuaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.313 - 322, 2004/12
This paper deals with a computer simulation of a helium-air counter flow in a rectangular channel. The inclination angle is varied from 0(horizontal) to 90(vertical). Velocity profiles and concentration profiles are calculated with a computer program VSOP sub-module. Following main features of the counter flow are discussed. (1) Time required to establish a quasi-steady state counter flow. (2) The relationship between the inclination angle and the flow patterns of the counter flow (3) The developing process of velocity profiles and concentration profiles (4) The relationship between the inclination angle of the channel and the velocity profiles of upwards flow and the downwards flow (5) The relationship between the concentration profile and the inclination angle (6) The relationship between the net in-flow rate and the inclination angle We compared the computed velocity profile and the net in-flow rate with experimental data. A good agreement is obtained between the calculation and the experiment.
Hattori, Hirofumi*; Sato, Hiroshi; Nagano, Yasutaka*
Nihon Kikai Gakkai Rombunshu, B, 70(696), p.1919 - 1926, 2004/08
no abstracts in English
Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki
Nihon Kikai Gakkai Rombunshu, B, 70(694), p.1518 - 1525, 2004/06
no abstracts in English
Takeda, Tetsuaki; Ohashi, Hirofumi; Inagaki, Yoshiyuki
Nihon Kikai Gakkai 2003-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.17 - 18, 2003/08
A technology development of a hydrogen production system by a nuclear heat are being performed as a heat application system of a high-temperature gas cooled reactor in the Japan Atomic Energy Research Institute. The objective of this study is to clarify heat transfer characteristics of the steam reformer in the HTTR hydrogen production system. An experiment has been performed using a double coaxial vertical tube to obtain the heat transfer characteristics and to evaluate the effectiveness of heat transfer enhancement. The amount of produced hydrogen increases with increasing not only reaction rate of catalysis but also the heat transfer coefficient. It is necessary to take into account of heat transfer from both surfaces of the double coaxial tube in order to obtain the amount of transferred heat from the heated tube to the coolant gas.
Ito, Toshimichi; Aramaki, Takafumi; Kitamura, Toshikatsu; Otosaka, Shigeyoshi; Suzuki, Takashi; Togawa, Orihiko; Kobayashi, Takuya; Senju, Tomoharu*; Chaykovskaya, E. L.*; Karasev, E. V.*; et al.
Journal of Environmental Radioactivity, 68(3), p.249 - 267, 2003/07
Times Cited Count：41 Percentile：62.78(Environmental Sciences)The anthropogenic radionuclides, 
Sr, 
Cs and 
Pu, in the seawater column of the Japan Sea were measured during 1997-2000. The vertical profiles of radionuclide concentrations showed their typical features; exponential decrease with depth for the Sr and Cs and surface minimum - subsurface maximum for the Pu, and there are no substantial differences between the present study and the previous ones. The area-averaged concentrations and the inventories of radionuclides in the Japan Sea are higher than those in the Northwest Pacific Ocean. In the spatial distributions, high inventory area extends and intrudes from the Japan Basin into the Yamato Basin. It is suggested that radionuclides sink by the vertical transport occurring mainly in the Japan Basin then advect into the Yamato Basin after detouring around the Yamato Rise, and finally, they are accumulated in the deep seawater of the Japan Sea.
Maruyama, Yu*; Moriyama, Kiyofumi; Nakamura, Hideo; Hirano, Masashi; Nakajima, K.*
Journal of Nuclear Science and Technology, 40(1), p.12 - 21, 2003/01
Times Cited Count：6 Percentile：40.97(Nuclear Science & Technology)no abstracts in English
Inaba, Yoshitomo; Takeda, Tetsuaki
JAERI-Research 2000-062, 73 Pages, 2001/02
JAERI-Research-2000-062.pdf:2.83MB
no abstracts in English
