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JAEA Reports

Summaries of research and development activities by using supercomputer system of JAEA in FY2017; April 1, 2017 - March 31, 2018

Information Technology Systems' Management and Operating Office

JAEA-Review 2018-018, 167 Pages, 2019/02

JAEA-Review-2018-018.pdf:34.23MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power and utilizes computational science and technology in many activities. As shown in the fact that about 20 percent of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2017, the system was used for R&D aiming to restore Fukushima (environmental recovery and nuclear installation decommissioning) as a priority issue, and for JAEA's major projects such as R&D of fast reactor cycle technology, research for safety improvement in the field of nuclear energy, and basic nuclear science and engineering research. This report presents a great number of R&D results accomplished by using the system in FY2017, as well as user support, operational records and overviews of the system, and so on.

Journal Articles

Multi-dimensional numerical investigation of sodium spray combustion; Benchmark analysis of SNL T3 experiment

Sonehara, Masateru; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Clark, A. J.*; Denman, M. R.*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 5 Pages, 2018/11

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

Effects of environmental factors inside the crevice on corrosion of stainless steel in high temperature water

Yamamoto, Masahiro; Sato, Tomonori; Igarashi, Takahiro; Ueno, Fumiyoshi; Soma, Yasutaka

Proceedings of European Corrosion Congress 2017 (EUROCORR 2017) and 20th ICC & Process Safety Congress 2017 (USB Flash Drive), 6 Pages, 2018/09

The authors have studied the differences between outer surface and the crevice-like portion of SUS316L in high pressurized and high temperature water containing dissolved oxygen. We have already introduced that changes in the characteristics of corrosion products along the crevice directions and gap width. It is suggested that the environmental conditions are different with the features of crevice from these results. In this report, we introduce the changes in oxide films with crevice gaps and comparison with the numerical simulation data utilizing of FEM calculation.

Journal Articles

Atmospheric modeling of $$^{137}$$Cs plumes from the Fukushima Daiichi Nuclear Power Plant; Evaluation of the model intercomparison data of the Science Council of Japan

Kitayama, Kyo*; Morino, Yu*; Takigawa, Masayuki*; Nakajima, Teruyuki*; Hayami, Hiroshi*; Nagai, Haruyasu; Terada, Hiroaki; Saito, Kazuo*; Shimbori, Toshiki*; Kajino, Mizuo*; et al.

Journal of Geophysical Research; Atmospheres, 123(14), p.7754 - 7770, 2018/07

We compared seven atmospheric transport model results for $$^{137}$$Cs released during the Fukushima Daiichi Nuclear Power Plant accident. All the results had been submitted for a model intercomparison project of the Science Council of Japan in 2014. We assessed model performance by comparing model results with observed hourly atmospheric concentrations of $$^{137}$$Cs, focusing on nine plumes over the Tohoku and Kanto regions. The results showed that model performance for $$^{137}$$Cs concentrations was highly variable among models and plumes. We also assessed model performance for accumulated $$^{137}$$Cs deposition. Simulated areas of high deposition were consistent with the plume pathways, though the models that best simulated $$^{137}$$Cs concentrations were different from those that best simulated deposition. The ensemble mean of all models consistently reproduced $$^{137}$$Cs concentrations and deposition well, suggesting that use of a multimodel ensemble results in more effective and consistent model performance.

Journal Articles

Some characteristics of gas-liquid two-phase flow in vertical large-diameter channels

Shen, X.*; Schlegel, J. P.*; Hibiki, Takashi*; Nakamura, Hideo

Nuclear Engineering and Design, 333, p.87 - 98, 2018/07

 Times Cited Count:1 Percentile:62.02(Nuclear Science & Technology)

Journal Articles

Three-dimensional numerical study on pool stratification behavior in molten corium-concrete interaction (MCCI) with MPS method

Li, X.; Sato, Ikken; Yamaji, Akifumi*; Duan, G.*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Molten corium-concrete interaction (MCCI) is an important ex-vessel phenomenon that could happen during the late phase of a hypothetical severe accident in a light water reactor. In the present study, a three-dimensional (3-D) numerical study has been performed to simulate COMET-L3 test carried out by KIT with a stratified molten pool configuration of simulant materials with improved MPS method. The heat transfer between corium/crust/concrete was modeled with heat conduction between particles. Moreover, the potential influence of the siliceous aggregates was also investigated by setting up two different case studies since there was previous study indicating that siliceous aggregates in siliceous concrete might contribute to different axial and radial concrete ablation rates. The simulation results have indicated that metal melt as corium in MCCI can have completely different characteristics regarding concrete ablation pattern from that of oxidic corium, which needs to be taken into consideration when assessing the containment melt-through time in severe accident management.

Journal Articles

Evaluation of melting and solidification processes by laser irradiations using a computational science simulation code SPLICE

Muramatsu, Toshiharu

Dai-89-Kai Reza Kako Gakkai Koen Rombunshu, p.115 - 119, 2018/05

no abstracts in English

JAEA Reports

Summaries of research and development activities by using supercomputer system of JAEA in FY2016; April 1, 2016 - March 31, 2017

Information Technology Systems' Management and Operating Office

JAEA-Review 2017-023, 157 Pages, 2018/02

JAEA-Review-2017-023.pdf:22.68MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. As shown in the fact that about 20% of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2016, the system was used for R&D aiming to restore Fukushima (environmental recovery and nuclear installation decommissioning) as a priority issue, as well as for JAEA's major projects such as research and development of fast reactor cycle technology, research for safety improvement in the field of nuclear energy, and basic nuclear science and engineering research. This report presents a great number of R&D results accomplished by using the system in FY2016, as well as user support, operational records and overviews of the system, and so on.

JAEA Reports

Development of a calculation method for atmospheric dispersion database that can immediately provide calculation results for any source term and period from hindcast to short-term forecast (Joint research)

Terada, Hiroaki; Tsuzuki, Katsunori; Kadowaki, Masanao; Nagai, Haruyasu; Tanaka, Atsunori*

JAEA-Data/Code 2017-013, 31 Pages, 2018/01

JAEA-Data-Code-2017-013.pdf:9.52MB

We developed an atmospheric dispersion calculation method that can respond to various needs for dispersion prediction in nuclear emergency and prepare database of information useful for planning of emergency response. In this method, it is possible to immediately get the prediction results for provided source term by creating a database of dispersion calculation results without specifying radionuclides, release rate and period except release point. By performing this calculation steadily along with meteorological data update, it is possible to immediately get calculation results for any source term and period from hindcast to short-term forecast. This function can be used for pre-accident planning such as optimization of monitoring plan and understanding events to be supposed for emergency response. Spatiotemporal distribution of radioactive materials reproduced by source term estimated inversely from monitoring based on this method is useful as a supplement to monitoring.

Journal Articles

Multi-dimensional gas-liquid two-phase flow in vertical large-diameter channels

Shen, X.*; Schlegel, J. P.*; Hibiki, Takashi*; Nakamura, Hideo

Proceedings of 2017 Japan-US Seminar on Two-Phase Flow Dynamics (JUS 2017), 6 Pages, 2017/06

Journal Articles

Development of thermal hydraulics analysis code ASFRE for fuel assembly of sodium-cooled fast reactor; Modification of distributed resistance model and validation analysis

Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.

JAEA Reports

Summaries of research and development activities by using supercomputer system of JAEA in FY2014; April 1, 2014 - March 31, 2015

Information Technology Systems' Management and Operating Office

JAEA-Review 2015-028, 229 Pages, 2016/02

JAEA-Review-2015-028.pdf:53.37MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. As shown in the fact that about 20% of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2014, the system was used. For R&D aiming to restore Fukushima (nuclear plant decommissioning and environmental restoration) as a priority issue, as well as for JAEA's major projects such as Fast Reactor Cycle System, Fusion R&D and Quantum Beam Science. This report presents a great amount of R&D results accomplished by using the system in FY2014, as well as user support, operational records and overviews of the system, and so on.

Journal Articles

Numerical simulations of gas-liquid-particle three-phase flows using a hybrid method

Guo, L.*; Morita, Koji*; Tobita, Yoshiharu

Journal of Nuclear Science and Technology, 53(2), p.271 - 280, 2016/02

 Times Cited Count:4 Percentile:31.55(Nuclear Science & Technology)

Journal Articles

Development of residual thermal stress-relieving structure of CFC monoblock target for JT-60SA divertor

Tsuru, Daigo; Sakurai, Shinji; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Suzuki, Satoshi

Fusion Engineering and Design, 98-99, p.1403 - 1406, 2015/10

 Times Cited Count:1 Percentile:81.41(Nuclear Science & Technology)

Journal Articles

Proposal of benchmark problem of thermal striping phenomena in planar triple parallel jets tests for fundamental code validation in sodium-cooled fast reactor development

Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08

Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.

Journal Articles

Numerical simulation of self-priming phenomena in venturi scrubber by two-phase flow simulation code TPFIT

Horiguchi, Naoki; Yoshida, Hiroyuki; Kanagawa, Tetsuya*; Kaneko, Akiko*; Abe, Yutaka*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

From the viewpoint of protecting containment and suppressing diffusion of the radioactive materials at severe accidents of nuclear power plant, it is important to install filtered venting devices to permit release of high pressure pollutant gas to the atmosphere by eliminating radioactive materials in the gas. A Multi Venturi Scrubber System (MVSS) is one of the devices for the filtered venting, and is used to realize filtered venting without any power supply. The MVSS is composed of a "Venturi Scrubbers" part and a "bubble column" part. In the Venturi Scrubbers part of the MVSS, there are hundreds of the Venturi scrubbers (VS). In an operation mode of the MVSS, the radioactive materials are eliminated through the gas-liquid interface from the pollutant gas to the liquid phase of a dispersed flow in the VS and a bubbly flow in the bubble column part. In the VS, the dispersed flow is formed from the liquid, which is suctioned through the hole for suction (called self-priming). In previous studies, an evaluation method to evaluate the liquid flow rate by the self-priming was developed. However, to develop evaluation methods of performance of the VSs, the two-phase flow behavior must be investigated, including droplet size and velocity difference of liquid and gas phases. Two-phase flow behavior in the VS is complicated, and it is difficult to estimate two-phase flow behavior of the VS by only experimental procedures. In this study, to investigate the hydraulic behavior of the VS, we tried to apply a detailed numerical simulation method of two-phase flow to the numerical simulation of the VS. In the simulation, TPFIT developed in JAEA was used as the detailed numerical simulation method. In this paper, we performed the numerical simulation air-water two-phase flow in the of the lab scale VS by the TPFIT, and numerical results were compared with experimental results.

JAEA Reports

Summaries of research and development activities by using supercomputer system of JAEA in FY2013; April 1, 2013 - March 31, 2014

Information Technology Systems' Management and Operating Office

JAEA-Review 2014-043, 241 Pages, 2015/02

JAEA-Review-2014-043.pdf:102.18MB

Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. About 20% of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology utilization. In FY2013, the system was used not only for JAEA's major projects such as Fast Reactor Cycle System, Fusion R&D and Quantum Beam Science, but also for R&D aiming to restore Fukushima (nuclear plant decommissioning and environmental restoration) as apriority issue. This report presents a great amount of R&D results accomplished by using the system in FY2013, as well as user support, operational records and overviews of the system, and so on.

Journal Articles

Current status of thermal/hydraulic feasibility project for reduced-moderation water reactor, 2; Development of two-phase flow simulation code with advanced interface tracking method

Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04

The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

Journal Articles

Incorporation of CO$$_{2}$$ exchange processes into a multilayer atmosphere-soil-vegetation model

Nagai, Haruyasu

Journal of Applied Meteorology, 44(10), p.1574 - 1592, 2005/10

This paper describes the incorporation of CO$$_{2}$$ exchange processes into an atmosphere-soil-vegetation model SOLVEG and examination of its sensitivity and impact of its stomatal resistance calculation on the latent heat flux over a winter wheat field. The model framework for the heat and water exchanges between the atmosphere and ground surface was validated in the previous papers (Nagai 2002, 2003). In this study, CO$$_{2}$$ exchange processes are incorporated in the model and the performance is examined. In the test calculation, the model simulated the CO$$_{2}$$ flux at 2 m above the ground well as a whole. A sensitivity test to clarify uncertainties for the model settings and parameters showed that the CO$$_{2}$$ production in the soil is the most important factor for the CO$$_{2}$$ calculation. Also, the impact of the CO$$_{2}$$ processes on the latent heat flux is discussed. The results indicate that the new model is effective and preferable to study surface exchanges of heat and water as well as CO$$_{2}$$.

121 (Records 1-20 displayed on this page)