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Journal Articles

Survey on technical issues of fission products behavior for improvement of decommissioning work efficiency and source term predicting accuracy; Report on the activity of this research committee for 2 years

Katsumura, Kosuke*; Takagi, Junichi*; Hosomi, Kenji*; Miyahara, Naoya*; Koma, Yoshikazu; Imoto, Jumpei; Karasawa, Hidetoshi; Miwa, Shuhei; Shiotsu, Hiroyuki; Hidaka, Akihide*; et al.

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 65(11), p.674 - 679, 2023/11

no abstracts in English

Journal Articles

Double diffusive dissolution model of UO$$_{2}$$ pellet in molten Zr cladding

Ito, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*

Journal of Nuclear Science and Technology, 60(4), p.450 - 459, 2023/04

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

Development of the simplified boiling model applied to the large-scale detailed simulation

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

Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09

Japan Atomic Energy Agency is developing the computational fluid dynamics code, JUPITER, based on the volume of fluid (VOF) method to analyze detailed thermal-hydraulics in a reactor. The detailed numerical simulation of boiling from a heating surface needs a substantial computational cost to resolve the microscale thermal-hydraulic phenomena such as the bubble generation from a cavity and evaporation of a micro-layer. This study developed the simplified boiling model from the heating surface to reduce the computational cost, which will apply to the detailed simulation code based on the surface tracking method such as JUPITER. We applied the simplified boiling model to JUPITER, and compared the simulation results with the experimental data of the vertical heating surface in the forced convection. We confirmed the degree of their reproducibility, and the issues to be modified were extracted.

Journal Articles

Numerical simulation of annular dispersed flow in simplified subchannel of light water cooled fast reactor RBWR

Yoshida, Hiroyuki; Horiguchi, Naoki; Ono, Ayako; Furuichi, Hajime*; Katono, Kenichi*

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08

Journal Articles

Development of the simplified boiling model applied for the large scale simulation by the detailed two-phase flow analysis based on the surface tracking

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

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07

JAEA is implementing a simulation of a two-phase flow in the reactor core by TPFIT and JUPITER which are developed by JAEA based on the surface tracking method. However, it is impossible to simulate a boiling on the heating surface in the large-scale domain by this type of simulation method since the simulation of boiling based on the surface tracking method needs the fine mesh which sufficiently resolves the initiation of boiling. Therefore, JAEA started to develop the simplified boiling model applied for the two-phase flow in the fuel assemblies. In this study, the simulation results of the convection boiling on a vertical heating surface and the comparison between the simulation results and experimental results are shown.

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

Experiments of melt jet-breakup for agglomerated debris formation using a metallic melt

Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Abe, Yutaka*

Nuclear Engineering and Design, 386, p.111575_1 - 111575_17, 2022/01

 Times Cited Count:3 Percentile:52.93(Nuclear Science & Technology)

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

Thermal-hydraulics to risk assessment; Roles of thermal-hydraulics simulation to risk assessment

Maruyama, Yu; Yoshida, Kazuo

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 63(7), p.517 - 522, 2021/07

no abstracts in English

Journal Articles

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

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

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

Journal Articles

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

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

Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

Miyahara, Naoya; Miwa, Shuhei; Horiguchi, Naoki; Sato, Isamu*; Osaka, Masahiko

Journal of Nuclear Science and Technology, 56(2), p.228 - 240, 2019/02

 Times Cited Count:8 Percentile:62.42(Nuclear Science & Technology)

In order to improve LWR source term under severe accident conditions, the first version of a fission product (FP) chemistry database named "ECUME" was developed. The ECUME is intended to include major chemical reactions and their effective kinetic constants for representative SA sequences. It is expected that the ECUME can serve as a fundamental basis from which FP chemical models in the SA analysis codes can be elaborated. The implemented chemical reactions in the first version were those for representative gas species in Cs-I-B-Mo-O-H system. The chemical reaction kinetic constants were evaluated from either literature data or calculated values using ab-initio calculations. The sample chemical reaction calculation using the presently constructed dataset showed meaningful kinetics effects at 1000 K. Comparison of the chemical equilibrium compositions by using the dataset with those by chemical equilibrium calculations has shown rather good consistency for the representative Cs-I-B-Mo-O-H species. From these results, it was concluded that the present dataset should be useful to evaluate FP chemistry in Cs-I-B-Mo-O-H system under LWA SA conditions.

JAEA Reports

Progress report on Nuclear Safety Research Center (JFY 2015 - 2017)

Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness

JAEA-Review 2018-022, 201 Pages, 2019/01

JAEA-Review-2018-022.pdf:20.61MB

Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

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

Technical basis of accident tolerant fuel updated under a Japanese R&D project

Yamashita, Shinichiro; Nagase, Fumihisa; Kurata, Masaki; Nozawa, Takashi; Watanabe, Seiichi*; Kirimura, Kazuki*; Kakiuchi, Kazuo*; Kondo, Takao*; Sakamoto, Kan*; Kusagaya, Kazuyuki*; et al.

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

In Japan, the research and development (R&D) project on accident tolerant fuel and other components (ATFs) of light water reactors (LWRs) has been initiated in 2015 for establishing technical basis of ATFs. The Japan Atomic Energy Agency (JAEA) has coordinated and carried out this ATF R&D project in cooperation with power plant providers, fuel venders and universities for making the best use of the experiences, knowledges in commercial uses of zirconium-base alloys (Zircaloy) in LWRs. ATF candidate materials under consideration in the project are FeCrAl steel strengthened by dispersion of fine oxide particles(FeCrAl-ODS) and silicon carbide (SiC) composite, and are expecting to endure severe accident conditions in the reactor core for a longer period of time than the Zircaloy while maintaining or improving fuel performance during normal operations. In this paper, the progresses of the R&D project are reported.

Journal Articles

Performance degradation of candidate accident-tolerant cladding under corrosive environment

Nagase, Fumihisa; Sakamoto, Kan*; Yamashita, Shinichiro

Corrosion Reviews, 35(3), p.129 - 140, 2017/08

 Times Cited Count:13 Percentile:50.67(Electrochemistry)

As the lessons learnt from the accident at the Fukushima Daiichi Nuclear Power Station, advanced cladding materials are being developed to enhance accident tolerance comparing with conventional zirconium alloys. The present paper reviews the progress of the development and summarizes subjects to be solved for the enhanced accident-tolerance fuel cladding, focusing on performance degradation under various corrosive environmental conditions that should be considered in designing the LWR fuel.

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