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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.
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; Sakashita, Hiroto*; Yoshida, Hiroyuki
Heat Transfer Engineering, 42(21), p.1775 - 1788, 2021/00
Times Cited Count：3 Percentile：22.73(Thermodynamics)In this study, the macrolayer formation model is proposed to predict the critical heat flux in the saturated and subcooled pool boiling based on the macrolayer dryout model. This model concept is based on the results of the previous experiments. In the model, the nucleation site is assumed to distribute based on the Poisson distribution. Combining the proposed macrolayer formation model and macrolayer dryout model, the CHFs up to subcooling 40K were predicted and they are successfully good agreement with the experimental data. Moreover, the concept of the model was confirmed by the numerical simulation using the TPFIT.
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.
Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10
The mechanism of Critical Heat Flux (CHF) remains to be clarified, even though it is important to evaluate the CHF for super high heat flux components such as light water reactors (LWRs). Some theoretical models to predict the CHF is proposed so far. A macrolayer formation model which is proposed in order to predict the CHF based on the macrolayer dryout model. In this model, it is assumed that the liquid is captured inside vapor mass at coalescence. In this study, the verification of the assumption of a macrolayer formation model by the numerical simulation of CMFD code, TPFIT, from the view point of hydrodynamics.
Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 6 Pages, 2018/07
The mechanism of critical heat flux (CHF) for higher system pressure remains to be clarified, even though it is important to evaluate the CHF for the light water reactor (LWR) which is operated under the high pressure condition. In this study, the process of bubble coalescence was simulated by using a computational multi-fluid dynamics (CMFD) simulation code TPFIT under various system pressure in order to investigate the behavior of bubbles as a basic study. The growth of bubbles was simulated by blowing of vapor from a tiny orifice simulating bubble bottom. One or four orifices were located on the bottom surface in this simulation study. The numerical simulations were conducted by varying the pressure and temperature.
Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao
JAERI-Review 2005-029, 119 Pages, 2005/09
JAERI-Review-2005-029.pdf:11.01MB
The research on Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed in JAERI for the development of future innovative reactors. The workshop on the FLWRs has been held every year since 1998 aiming at information exchange between JAERI and other organizations. The 8th workshop was held on Feb. 10, 2005 under the joint auspices of JAERI and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 75 participants. The workshop began with 3 presentations on FLWRs entitled "Framework and Status of Research and Development on FLWRs", "Long-Term Fuel Cycle Scenarios for Advanced Utilization of Plutonium from LWRs", and "Experiments on Characteristics on Hydrodynamics in Tight-Lattice Core". Then 3 lectures followed: "Development of Evaluation Method for Accuracy in Predicting Neutronics Characteristics of Tight-Lattice Core" by Osaka University, "Development of Cost-Reduced Low-Moderation Spectrum Boiling Water Reactor" by Toshiba Corporation and "Design and Analysis on Super-Critical Water Cooled Power Reactors" by Tokyo University.
Liu, W.; Nariai, Hideki*
Journal of Heat Transfer, 127(2), p.149 - 158, 2005/02
Times Cited Count：12 Percentile：45.93(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.
Liu, W.; Kureta, Masatoshi; Onuki, Akira; Akimoto, Hajime
Journal of Nuclear Science and Technology, 42(1), p.40 - 49, 2005/01
Times Cited Count：7 Percentile：44.84(Nuclear Science & Technology)In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For low mass velocity region (
300 kg/m
s), the correlation is written in critical quality - annular flow length type. For high mass velocity region (
300 kg/ms), it is written in local critical heat flux - critical quality type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: gap between rods from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2000 kg/ms and pressure from 2 to 11 MPa.
Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato; Ibbott, C.*; Tivey, R.*
Fusion Science and Technology, 46(4), p.521 - 529, 2004/12
Times Cited Count：8 Percentile：48.76(Nuclear Science & Technology)Thermal hydraulic tests measuring critical heat flux CHF and pressure drop of an annular tube with twisted fin, "annular swirl tube", have been. This tube consists of two concentric tubes, the outer tube and the inner tube with a twisted fin on its outer surface. Cooling water flows inside of the inner tube first, and then returns into an annulus with a swirl flow at an end-return of the cooling tube. The CHF testing shows the no degradation of CHF of the annular swirl tube in comparison with the conventional swirl tube. A minimum axial velocity of 7.1m/sec is required for 28MW/m, the ITER design value. Applicability of the JAERI's correlation for the heat transfer to the annular swirl tube is also demonstrated by the comparison of the experimental results with those of the numerical analyses. The friction factor correlation for the annular flow with the twisted fins is made for the hydraulic designing of the vertical target. The least pressure drop at the end-return is obtained by using the hemispherical end-plug. Its radius is the same as that of ID of the outer cooling tube.
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
Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato
JAERI-Tech 2003-084, 49 Pages, 2003/11
JAERI-Tech-2003-084.pdf:1.94MB
Thermal hydraulic tests measuring critical heat flux CHF and pressure drop of an annular tube with twisted fin, "annular swirl tube", have been. This tube consists of two concentric tubes, the outer tube and the inner tube with a twisted fin on its outer surface. Cooling water flows inside of the inner tube first, and then returns into an annulus with a swirl flow at an end-return of the cooling tube. The CHF testing shows the no degradation of CHF of the annular swirl tube in comparison with the conventional swirl tube. A minimum axial velocity of 7.1m/sec is required for 28MW/m, the ITER design value. Applicability of the JAERI's correlation for the heat transfer to the annular swirl tube is also demonstrated by the comparison of the experimental results with those of the numerical analyses. The friction factor correlation for the annular flow with the twisted fins is made for the hydraulic designing of the vertical target. The least pressure drop at the end-return is obtained by using the hemispherical end-plug. Its radius is the same as that of ID of the outer cooling tube.
Nabeshima, Kunihiko; Nakatsuka, Toru; Ishikawa, Nobuyuki; Uchikawa, Sadao
JAERI-Conf 2003-020, 240 Pages, 2003/11
JAERI-Conf-2003-020.pdf:27.66MB
The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since 1998 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The workshop began with five lectures on status of research and development on RMWRs in JAERI entitled "Status and Future Program of Research and Development on Reduced-Moderation Water Reactors", "Design of Small Reduced-Moderation Water Reactors", "Critical Experiments for Reduced-Moderation Water Reactors", "Critical Heat Flux Experiments in Tight Lattice Core" and "Development of High Performance Cladding". Then two lectures followed: "Status of Phase II of Feasibility Studies on Commercialized Fast Breeder Reactor System" by JNC and "Present Status of Study on Super-critical water Cooled Power Reactor" by Toshiba Corporation.
Nakatsuka, Toru; Tamai, Hidesada; Kureta, Masatoshi; Okubo, Tsutomu; Akimoto, Hajime; Iwamura, Takamichi
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 6 Pages, 2003/09
It is important to evaluate thermal margin of the tight lattice core in the Reduced-Moderation Water reactor (RMWR). In the present study, to assess the applicability of subchannel analysis for tight lattice cores, tight lattice CHF experiments were analyzed with COBRA-TF code. For the axial uniform heated rod bundle, the code gives good prediction of critical power for mass velocity of around 500kg/(ms), while the code underestimates it for lower mass velocity and overestimates for higher mass velocity. The predicted BT position was outer channels and differed from the measured position. For the axially double-humped heated bundle, the code gives good prediction for mass velocity of around 200kg/(ms), and overestimates for higher mass velocity. It turned out that the two-phase multiplier of friction loss have a large influences on the flow distribution among the subchannels. To improve the calculation accuracy, it is required to predict precisely the flow distribution including the prediction of pressure distribution in a tight lattice bundle.
Akiba, Masato; Ishitsuka, Etsuo; Enoeda, Mikio; Nishitani, Takeo; Konishi, Satoshi
Purazuma, Kaku Yugo Gakkai-Shi, 79(9), p.929 - 934, 2003/09
no abstracts in English
Mishima, Kaichiro*; Saito, Yasushi*
JAERI-Tech 2002-014, 83 Pages, 2002/03
JAERI-Tech-2002-014.pdf:6.83MB
no abstracts in English
Ezato, Koichiro; Suzuki, Satoshi; Sato, Kazuyoshi; Taniguchi, Masaki; Hanada, Masaya; Araki, Masanori; Akiba, Masato
Fusion Engineering and Design, 56-57, p.291 - 295, 2001/10
Times Cited Count：14 Percentile：68.91(Nuclear Science & Technology)no abstracts in English
Kinoshita, Hidetaka; Nariai, Hideki*; Inasaka, Fujio*
JSME International Journal, Series B, 44(1), p.81 - 89, 2001/01
no abstracts in English
Akiba, Masato; Ezato, Koichiro; Sato, Kazuyoshi; Suzuki, Satoshi; Hatano, Toshihisa
Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering (SOFE '99), p.381 - 384, 1999/10
no abstracts in English
J.Boscary*; Araki, Masanori; ; *; Akiba, Masato
Fusion Technology, 35(3), p.289 - 296, 1999/05
no abstracts in English
