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Journal Articles

CFD analysis of natural circulation in LBE-cooled accelerator-driven system

Sugawara, Takanori; Watanabe, Nao; Ono, Ayako; Nishihara, Kenji; Ichihara, Kyoko*; Hanzawa, Kohei*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 10 Pages, 2022/03

Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides (MAs) included in high level wastes discharged from nuclear power plants. The ADS is a lead-bismuth cooled tank-type reactor with 800 MW thermal power. It is supposed that the ADS is safer than conventional critical reactors because it is operated in a subcritical state. The previous study performed the transient analyses for the typical ADS accidents such as unprotected loss of flow or beam overpower. It was shown that all calculation cases except loss of heat sink (LOHS) satisfied the no-damage criteria. To avoid the damage by LOHS, the ADS equips Direct Reactor Auxiliary Cooling System (DRACS) to remove the decay heat. The most important points of a DRACS operation are its reliability and to ensure the flowrate in a natural circulation state. This study aims to perform the CFD analysis of the natural circulation to clarify the flowrate in the ADS reactor vessel.

Journal Articles

Numerical simulation of two-phase flow in fuel assemblies with a spacer grid using a mechanistically based method

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.

Journal Articles

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$$Sigma$$, 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.

Journal Articles

Summary results of subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))"

Koyama, Shinichi; Nakagiri, Toshio; Osaka, Masahiko; Yoshida, Hiroyuki; Kurata, Masaki; Ikeuchi, Hirotomo; Maeda, Koji; Sasaki, Shinji; Onishi, Takashi; Takano, Masahide; et al.

Hairo, Osensui Taisaku jigyo jimukyoku Homu Peji (Internet), 144 Pages, 2021/08

JAEA performed the subsidy program for the "Project of Decommissioning and Contaminated Water Management (Development of Analysis and Estimation Technology for Characterization of Fuel Debris (Development of Technologies for Enhanced Analysis Accuracy and Thermal Behavior Estimation of Fuel Debris))" in 2020JFY. This presentation summarized briefly the results of the project, which will be available shortly on the website of Management Office for the Project of Decommissioning and Contaminated Water Management.

Journal Articles

Core thermal-hydraulic analysis during dipped-type direct heat exchanger operation in natural circulation conditions

Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*

Proceedings of 28th International Conference on Nuclear Engineering; Nuclear Energy the Future Zero Carbon Power (ICONE 28) (Internet), 10 Pages, 2021/08

To enhance the safety of sodium-cooled fast reactors, a dipped-type direct heat exchanger (D-DHX) has been investigated in a natural circulation decay heat removal system. During the D-DHX operation, the core-plenum interactions occurs, therefore, a thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics code (RV-CFD model) is necessarily required. In this study, the application of the subchannel analysis method for subassemblies to the RV-CFD model was attempted to reduce the calculation costs. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1. As the result, the behavior of cold sodium into the simulated core was well grasped and the calculated sodium temperature in the core had good agreement with the experimental result. The applicability of the RV-CFD model was confirmed.

Journal Articles

Macrolayer formation model for prediction of critical heat flux in saturated and subcooled pool boiling

Ono, Ayako; Sakashita, Hiroto*; Yoshida, Hiroyuki

Heat Transfer Engineering, 42(21), p.1775 - 1788, 2021/00

 Times Cited Count:0 Percentile:0.01(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.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

Journal Articles

Numerical simulation of two-phase flow in 4$$times$$4 simulated bundle

Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki

Mechanical Engineering Journal (Internet), 7(3), p.19-00583_1 - 19-00583_12, 2020/06

JAEA is implementing the 3D detailed nuclear-thermal-coupled analysis code to analyze the transition state of the core and to reduce the likelihood of the design. In the development plan, the computational fluid dynamics code based on the VOF method, JUPITER, is applied for TH part of the 3D detailed nuclear-thermal-coupled analysis code.

Journal Articles

Sea water flow boiling heat transfer involving sea salt deposition; Role of deposited sea salt

Koizumi, Yasuo*; Uesawa, Shinichiro; Ono, Ayako; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Nihon Kikai Gakkai Netsu Kogaku Konfarensu 2019 Koen Rombunshu (USB Flash Drive), 1 Pages, 2019/10

no abstracts in English

Journal Articles

Study on the two-phase flow in simulated LWR fuel bundle by CFD code

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$$times$$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.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. In this paper, the numerical simulation results of the preliminary analysis for the sodium experiment with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX.

Journal Articles

Measurement of void fraction distribution in a 4$$times$$4 fuel bundle under high pressure condition for validation of two-phase CFD code

Nagatake, Taku; Shibata, Mitsuhiko; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

In the Fukushima Daiichi Nuclear Power Plant accident, reactor cores were cooled by natural circulation due to pump trip. To investigate the accident progress of the Fukushima Daiichi Nuclear Power Plant, it is important to understand the thermal hydraulic behavior in reactor cores including fuel bundles. Flow rate inside cores was relatively low in the natural circulation conditions, then, thermal-hydraulic behavior in the fuel bundles was different from that in the normal operating conditions. To evaluate thermal hydraulic behavior under the accidental conditions, we are developing the numerical simulation codes named TPFIT and ACE3D. These codes are based on two-phase computational fluid dynamics and can simulate the two-phase flow inside fuel bundles including low flow rate condition. Before applying these codes to the thermal-hydraulic behavior, the applicability of these codes must be confirmed. Then, in this study, in order to obtain a validation data for TPFIT and ACE3D code, thermal hydraulic experiment was performed by using test section with a simulated fuel bundle with 4$$times$$4 unheated rods. In this simulated fuel bundle, there were wire mesh sensors, and void fraction distribution data inside the simulated fuel bundle under high pressure condition (max. 2.6 MPa) was obtained. The one of the advantage of wire mesh sensor is that a void fraction distribution of cross section at the same time can be measured. In this paper, void fraction distribution of two-phase flow in a simulated fuel bundle under high pressure condition are reported.

Journal Articles

Simultaneous measurement of dry patch behavior and surface temperature distribution for copper heat transfer surface with deposition layer in nucleate pool boiling

Uesawa, Shinichiro; Ono, Ayako; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Nihon Kikai Gakkai Netsu Kogaku Konfarensu 2018 Koen Rombunshu (USB Flash Drive), 6 Pages, 2018/10

no abstracts in English

Journal Articles

Numerical study on effect of nucleation site density on behavior of bubble coalescence by using CMFD simulation code TPFIT

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.

Journal Articles

Development of numerical estimation method for thermal hydraulics in reactor vessel of sodium-cooled fast reactor under decay heat removal system operation conditions; Preliminary thermal hydraulics simulation for simulated reactor vessel in sodium experimental apparatus PLANDTL-2

Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).

Journal Articles

Numerical study on effect of pressure on behavior of bubble coalescence by using CMFD simulation

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.

Journal Articles

Observation of dry patch behavior on copper heat transfer surface and measurement of surface temperature distribution in nucleate pool boiling

Uesawa, Shinichiro; Ono, Ayako; Koizumi, Yasuo; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Dai-55-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 8 Pages, 2018/05

no abstracts in English

Journal Articles

Study on applicability of fast reactor plant dynamics analysis code to core thermal hydraulics under natural circulation decay heat removal conditions

Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04

A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.

Journal Articles

Study on reactor vessel coolability of sodium-cooled fast reactor under severe accident condition; Water experiments using a scale model

Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Kamide, Hideki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

The water experiment apparatus simulating the thermal hydraulics in a reactor vessel under operating the decay heat removal systems (DHRSs) was fabricated. The theoretical evaluation for similarity and results of basic experiments show applicability for a scale model experiment of a sodium-cooled fast reactor. This paper, moreover, describes the results of flow visualization experiment under operating a dipped-type passive DHX, which is planned to be installed in both a loop type reactor and pool type reactor, and the calculation results using FLUENT comparing with the result of water experiment.

Journal Articles

Influence of inlet velocity condition on unsteady flow characteristics in piping with a short elbow under a high-Reynolds-number condition

Ono, Ayako; Tanaka, Masaaki; Kobayashi, Jun; Kamide, Hideki

Mechanical Engineering Journal (Internet), 4(1), p.16-00217_1 - 16-00217_15, 2017/02

In the design of the Advanced Sodium-cooled Fast Reactor in Japan, the Reynolds number in the primary hot leg (H/L) piping reaches 4.2$$times$$10$$^{7}$$. Furthermore, a short elbow is used in the H/L piping to achieve a compact plant layout. In the H/L piping, flow-induced vibration is a concern due to the excitation force caused by pressure fluctuation in the short elbow. In this report, the influence of inlet velocity condition on the unsteady velocity characteristics in the short elbow was studied by controlling the flow patterns at the elbow inlet. Measured velocity distributions indicated that the inlet velocity profiles affected a circumferential secondary flow, which then affected an area of flow separation at the elbow. It was also found that the velocity fluctuation at low frequency components observed upstream of the elbow could remain in downstream of the elbow though its intensity was attenuated.

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