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Journal Articles

Numerical simulation of heat transfer behavior in EAGLE ID1 in-pile test using finite volume particle method

Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji

Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Sodium fire collaborative study progress; CNWG fiscal year 2020

Louie, D. L. Y.*; Aoyagi, Mitsuhiro

SAND2021-0136 (Internet), 53 Pages, 2021/01

This report discusses the progress on the collaboration between Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) on the sodium fire research in fiscal year 2020. First, the current sodium pool fire model in MELCOR is discussed. The associated sodium fire input requirements are also presented. These input requirements are flexible enough to permit further model development via control functions to enhance the current model without modifying the source code. The theoretical pool fire model improvement developed at SNL is discussed. A control function model has been developed from this improvement. Then, the validation study of the sodium pool fire model in MELCOR is described. To validate this pool fire model with the enhancement, a JAEA F7-1 sodium pool fire experiment is used. The results of the calculation are discussed as well as suggestions for further model improvement. Finally, recommendations are made for new MELCOR simulations for next fiscal year, 2021.

JAEA Reports

Upgrading of recovery method for radioactive microparticles by heavy liquid separation aiming to volume reduction of contaminated soil (Contract research); FY2019 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; University of Tsukuba*

JAEA-Review 2020-037, 53 Pages, 2020/12

JAEA-Review-2020-037.pdf:3.46MB

JAEA/CLADS had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project in FY2019. Among the adopted proposals in FY2018, this report summarizes the research results of the "Upgrading of Recovery Method for Radioactive Microparticles by Heavy Liquid Separation Aiming to Volume Reduction of Contaminated Soil" conducted in FY2019.

Journal Articles

Experiments of self-wastage phenomena elucidation in steam generator tube of sodium-cooled fast reactor

Umeda, Ryota; Shimoyama, Kazuhito; Kurihara, Akikazu

Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(4), p.234 - 244, 2020/12

Sodium-water reaction caused by failure of the steam generator tube of sodium-cooled fast reactor induce the wastage phenomenon, which has erosive and corrosive feature. In this report, the authors have performed the self-wastage experiments under high sodium temperature condition to evaluate the effect of wastage form/geometry by using two types of initial defect such as the micro fine pinhole and fatigue crack, and water leak rate on self-wastage rate. Based on the consideration of crack type influence, it was confirmed that self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide intervenes and inhibits the progress of self-wastage. The dependence of initial sodium temperature on self-wastage rate was clearly observed, and new self-wastage correlation was derived considering the initial sodium temperature.

Journal Articles

Validation of analysis models on relocation behavior of molten core materials in sodium-cooled fast reactors based on the melt discharge experiment

Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

Journal Articles

Analytical study on removal mechanisms of cesium aerosol from a noble gas bubble rising through liquid sodium pool

Miyahara, Shinya*; Kawaguchi, Munemichi; Seino, Hiroshi

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

In a postulated accident of fuel pin failure of sodium cooled fast reactor, a fission product cesium will be released as an aerosol such as cesium iodide and/or oxide together with xenon and/or krypton. In this study, cesium aerosol removal behavior due to inertial deposition, sedimentation and diffusion was analyzed by a computer program which deals with the expansion and the deformation of the bubble together with the aerosol absorption. Initial bubble diameter, sodium pool depth and temperature, aerosol particle diameter and density, initial aerosol concentration were changed as parameter. From the results, it was concluded that the initial bubble diameter was most sensitive parameter to the decontamination factor (DF). It was found that the sodium pool depth, the aerosol particle diameter and density have also important effect on the DF, but the sodium temperature has a marginal effect. To meet these results, the experiments are under planning to validate the results.

Journal Articles

Development of ex-vessel phenomena analysis model for multi-scenario simulation system, spectra

Uchibori, Akihiro; Aoyagi, Mitsuhiro; Takata, Takashi; Ohshima, Hiroyuki

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08

The multi-scenario simulation system named SPECTRA has been developed for integrated analysis of in- and ex-vessel phenomena during a severe accident in sodium-cooled fast reactors. The base module computing ex-vessel compressible gas behavior by a lumped mass model and a sodium-concrete interaction module were verified through the basic analyses individually. A validity of the system including the base module and the individual physical module such as the sodium-concrete interaction module was confirmed through the analysis assuming sodium leakage from a reactor vessel and a primary cooling loop.

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

Advancement of elemental analytical model in LEAP-III code for tube failure propagation

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Li, J.*; Jang, S.*

Mechanical Engineering Journal (Internet), 7(3), p.19-00548_1 - 19-00548_11, 2020/06

Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.

Journal Articles

A Conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchita, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.

Mechanical Engineering Journal (Internet), 7(3), p.19-00489_1 - 19-00489_16, 2020/06

The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.

Journal Articles

Influence of mixing solution on characteristics of calcium aluminate cement modified with sodium polyphosphate

Irisawa, Keita; Garcia-Lodeiro, I.*; Kinoshita, Hajime*

Cement and Concrete Research, 128, p.105951_1 - 105951_7, 2020/02

 Times Cited Count:2 Percentile:54.21(Construction & Building Technology)

This study investigated characteristics of a calcium aluminate cement modified with a phosphate (CAP) by changing an amount and concentration of mixing solution with sodium polyphosphate. When the amount of mixing solution was increased with a constant amount of sodium polyphosphate, an enhanced consumption of monocalcium aluminate was observed compared with gehlenite in calcium aluminate cement (CAC). Formation of gibbsite, Al(OH)$$_{3}$$, was also increased as a hydration product in the CAP and the possible reduction of water in the amorphous gel phase. When the amount of mixing solution was increased with a constant concentration of sodium polyphosphate, the enhanced consumption of monocalcium aluminate was not observed. Neither gibbsite nor any other crystalline hydration products were identified in this series. In addition, unreacted sodium polyphosphate remained in the system. The increased formation of gibbsite and the possible reduction of water from the amorphous gel phase appears to contribute to the improvement of the microstructure in the products.

JAEA Reports

Upgrading of recovery method for radioactive microparticles by heavy liquid separation aiming to volume reduction of contaminated soil (Contract research); FY2018 Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development

Collaborative Laboratories for Advanced Decommissioning Science; University of Tsukuba*

JAEA-Review 2019-023, 33 Pages, 2020/01

JAEA-Review-2019-023.pdf:1.97MB

CLADS, JAEA, had been conducting the Center of World Intelligence Project for Nuclear Science/Technology and Human Resource Development (hereafter referred to "the Project") in FY2018. The Project aims to contribute to solving problems in nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2018, this report summarizes the research results of the 'Upgrading of Recovery Method for Radioactive Microparticles by Heavy Liquid Separation Aiming to Volume Reduction of Contaminated Soil'. After the accident of the Fukushima Daiichi Nuclear Power Station, radioactive cesium has been heterogeneously distributed in surface soil due to the existence of radioactive microparticles and clay minerals. Therefore, the selective removal of these microparticles will lead to the volume reduction of contaminated soil. The present study examines methods for selectively removing radioactive microparticles from soil. Also, in order to reduce the volume of contaminated soil, we search a possibility to practically apply the separation method that uses the difference in specific gravity of particles (heavy liquid separation method).

Journal Articles

Sodium fire collaborative study progress; CNWG Fiscal Year 2019

Louie, D. L. Y.*; Uchibori, Akihiro

SAND2019-15043 (Internet), 35 Pages, 2019/12

This report describes the progress on the sodium fire research in fiscal year 2019 in the Civil Nuclear Energy Research and Development Working Group (CNWG). In this study, the validation study of the sodium pool fire model incorporated into the MELCOR code, which was originally developed for accident analysis in light water reactors, was carried out through the numerical analysis on the sodium pool fire experiment named F7-1. In this preliminary analysis, pool and atmosphere temperature went up to the same level with the measured results, while the unnatural behavior appeared in the latter half of the analysis. Based on this result, recommendations for improvement were made for a new analysis in next fiscal year, 2020.

Journal Articles

Study on analysis method for inert gas behavior in liquid metal flow with considering dissolution and entrainment at free surface

Matsushita, Kentaro; Ito, Kei*; Ezure, Toshiki; Tanaka, Masaaki

Dai-24-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2019/06

In the design study on a sodium-cooled fast reactor (SFR), a numerical simulation code named SYRENA has been developed in Japan Atomic Energy Agency to analyze the behavior of gas bubbles and/or dissolved gas in the primary coolant system. In the present study, the effect of the non-condensable gas entrainment at the free surface on the bubble and the dissolved gas behavior in the primary coolant system were investigated for a typical pool type reactor, and also effect of a dipped-plate (D/P) installed below the free surface in the reactor vessel to suppress the gas bubble entrainment into the primary coolant system was especially investigated. It was clarified that the D/P was influential to the non-condensable gas behavior and the molar flow rate of gas bubbles in the primary coolant system varies depending on the relationship between the gas entrainment rate at the free surface and the exchange flow rate through the D/P.

Journal Articles

Visualizing an ignition process of hydrogen jets containing sodium mist by high-speed imaging

Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 56(6), p.521 - 532, 2019/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Routing study of above core structure with mock-up experiment for ASTRID

Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.

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

Parametric analysis of bubble and dissolved gas behavior in primary coolant system of sodium-cooled fast reactors

Matsushita, Kentaro; Ito, Kei*; Ezure, Toshiki; Tanaka, Masaaki

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

A numerical simulation code named SYRENA has been developed in JAEA to analyze the behavior of entrained bubbles and dissolved gas in the primary coolant of sodium-cooled fast reactor (SFR). In the present study, a flow network model of SYRENA to a hypothetical pool type reactor was developed and the non-condensable gas behavior was investigated through the comparison with that in the loop type reactor. The effect of the dipped-plate (D/P) tentatively introduced into the pool-type reactor on the gas behavior was investigated through the parametric analyses about the sodium exchange flow rate through the D/P and the gas entrainment rate at the free surface. It was suggested that the increase in the exchange flow rate through the D/P doesn't always work to decrease the bubble volume in the primary coolant system.

Journal Articles

Multi-dimensional numerical benchmark analysis of SNL T3 sodium spray combustion experiment with AQUA-SF code

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

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In order to investigate the effect of sodium combustion, Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) have exchanged information of sodium combustion modelling and related experimental data in the framework of Civil Nuclear Energy Research and Development Working Group (CNWG). The benchmark analysis of the SNL T3 sodium spray combustion experiment and sensitivity study have been carried out using the AQUA-SF code in this paper. The sensitivity analysis clarifies the influencing factors of the multi-dimensional analysis such as turbulence models, radiation heat transfer model from sodium droplets, and momentum exchange between gas and droplets. The result shows that the turbulence effect, radiation from droplets and gas temperature increase at spray burning area affect sodium spray burning rate significantly.

Journal Articles

A Conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchita, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.

182 (Records 1-20 displayed on this page)