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Journal Articles

Pressure resistance thickness of disposal containers for spent fuel direct disposal

Sugita, Yutaka; Taniguchi, Naoki; Makino, Hitoshi; Kanamaru, Shinichiro*; Okumura, Taisei*

Nippon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.121 - 135, 2020/09

A series of structural analysis of disposal containers for direct disposal of spent fuel was carried out to provide preliminary estimates of the required pressure resistance thickness of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body and then the lid of the disposal container. This work also provides additional analytical technical knowledge, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.

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.

JAEA Reports

Summaries of research and development activities by using supercomputer system of JAEA in FY2018 (April 1, 2018 - March 31, 2019)

HPC Technology Promotion Office*

JAEA-Review 2019-017, 182 Pages, 2020/01

JAEA-Review-2019-017.pdf:11.11MB

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 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 FY2018, 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 FY2018, as well as user support, operational records and overviews of the system, and so on.

Journal Articles

Effects of heterogeneity of geomechanical properties on tunnel support stress during tunnel excavation

Okazaki, Yasuyuki*; Hayashi, Hisashi*; Aoyagi, Kazuhei; Morimoto, Shingo*; Shinji, Masato*

Proceedings of 5th ISRM Young Scholars' Symposium on Rock Mechanics and International Symposium on Rock Engineering for Innovative Future (YSRM 2019 and REIF 2019) (USB Flash Drive), 6 Pages, 2019/12

In the design of tunnel support, the behavior of the rock mass around a tunnel and the stress acting on the tunnel support may be predicted using a numerical analysis. However, in such a numerical analysis, it is common to assume that each stratum comprises a homogeneous material, ignoring the heterogeneity of the geomechanical properties inherent to the rock mass. For this reason, it is not unusual for the results of the numerical analysis to differ from the actual behavior. We performed a tunnel excavation analysis considering the heterogeneity of the geomechanical properties in the rock mass to investigate the local increase in the tunnel support stress obtained in the 350 m gallery at the Horonobe Underground Research Laboratory. The results revealed that, in order to predict the locally increased support stress in advance, it is necessary to carry out a tunneling excavation analysis considering the heterogeneity of the geomechanical properties. It was also revealed that the scale at which the geomechanical properties fluctuate is an important factor.

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

Development of numerical analysis method for tube failure propagation under sodium-water reaction accident

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11

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.

Journal Articles

Advancement of numerical analysis method for tube failure propagation

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

Proceedings of 2018 ANS Winter Meeting and Nuclear Technology Expo; Embedded Topical International Topical Meeting on Advances in Thermal Hydraulics (ATH 2018) (USB Flash Drive), p.1289 - 1294, 2018/11

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

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.

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.

Journal Articles

A Simple and practical correction technique for reactivity worth of short-sized samples measured by critical-water-level method

Kitamura, Yasunori*; Fukushima, Masahiro

Nuclear Science and Engineering, 186(2), p.168 - 179, 2017/05

 Times Cited Count:1 Percentile:79.72(Nuclear Science & Technology)

An inconsistency between the reactivity worth of short-size samples measured by the critical-water-level (CWL) method and that conventionally analysed for validating the nuclear data and the nuclear calculation methods has been known. The present study investigated this inconsistency in terms of a simple theoretical framework and proposed a simple and practical technique for correcting the measured sample reactivity worth without making supplementary experiments. A series of Monte Carlo calculations that simulated typical sample reactivity worth measurement by the CWL method showed that this inconsistency is effectively reduced by the present correction technique.

Journal Articles

Study on self-wastage phenomenon at heat transfer tube in steam generator of sodium-cooled fast reactor with consideration of thermal coupling of fluid and structure

Kojima, Saori*; Uchibori, Akihiro; Takata, Takashi; Ohno, Shuji; Fukuda, Takeshi*; Yamaguchi, Akira*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 8 Pages, 2016/11

Analytical evaluation on a self-wastage phenomenon at heat transfer tubes in the steam generator of sodium cooled fast reactors has been performed by using the sodium-water reaction analysis code SERAPHIM. In this study, a fluid-structure thermal coupling model was developed and incorporated in the SERAPHIM code to evaluate heat transfer between the sodium-side reacting flow and the outer surface of the heat transfer tube. The effect of the fluid-structure thermal coupling model on the temperature field was demonstrated through the numerical analyses.

Journal Articles

Development of unstructured mesh-based numerical method for sodium-water reaction phenomenon in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki

Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-6) (Internet), 11 Pages, 2016/09

For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an under expanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.

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

The Effect of the inhomogeneity of rock mass to the tunnel support stress during tunnel excavation

Okazaki, Yasuyuki*; Aoyagi, Kazuhei; Kumasaka, Hiroo*; Shinji, Masato*

Doboku Gakkai Rombunshu, F1 (Tonneru Kogaku) (Internet), 72(3), p.I_1 - I_15, 2016/00

In the rational tunnel support design, numerical analysis is powerful tool to know the estimation of the behavior before tunnel construction in the case of the special ground condition and limited similar construction. In order to evaluate the support structure quantitatively, it is necessary to understand the effect of the inhomogeneity of rock mass to the tunnel support stress in advance. In this study, tunnel excavation analysis considering the inhomogeneity of rock mass was carried out. The analysis results were compared with the stress measured in the tunnel support in the Horonobe underground research laboratory. As a result, it was revealed that the local stress measured in the tunnel support can be simulated by considering the inhomogeneity of rock mass stochastically. In addition, this study evaluated the effect of the inhomogeneity of rock mass to the tunnel support stress quantitatively by processing analysis results statistically.

Journal Articles

Development of a multiphysics analysis system for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

AIP Conference Proceedings 1702, p.040010_1 - 040010_4, 2015/12

 Times Cited Count:0 Percentile:100

A Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence in a steam generator of sodium-cooled fast reactors. The system consists of the computer codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow involving sodium-water chemical reaction. Applicability of the SERAPHIM code was confirmed through the analyses of the basic experiment and the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models of the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the rapidly heated tube. The developed system enabled us to evaluate the wastage environment and possibility of failure propagation.

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

CFD analysis

Takase, Kazuyuki; Misawa, Takeharu*

Supercritical-Pressure Light Water Cooled Reactors, p.301 - 319, 2014/12

no abstracts in English

Journal Articles

A Study of mechanical stability of support elements and surrounding rock mass during shaft sinking through a fault

Tsusaka, Kimikazu*; Inagaki, Daisuke*; Nago, Makito*; Ijiri, Yuji*

Proceedings of 8th Asian Rock Mechanics Symposium (ARMS-8) (USB Flash Drive), 9 Pages, 2014/10

Journal Articles

Large-scale direct simulation of two-phase flow structure around a spacer in a tight-lattice nuclear fuel bundle

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime

Computational Fluid Dynamics 2004, p.649 - 654, 2006/00

no abstracts in English

78 (Records 1-20 displayed on this page)