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4 simulated fuel bundle for validation of thermal-hydraulics simulation codesNagatake, Taku; Shibata, Mitsuhiko; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
JAEA is developing a neutronics/thermal-hydraulics coupling simulation code for light-water reactors. Thermal-hydraulic simulation codes applied to the platform are expected to evaluate void fraction distributions in fuel assemblies under operational conditions, which is necessary for neutron transport simulation, and need to be validated using void fraction distribution data in a rod bundle under high-temperature and high-pressure conditions. To obtain the data for code validation, we have been measuring the instantaneous void fraction distribution in a 44 simulated fuel assembly by a wire mesh sensor. In this paper, we report the results of the experiments with pressure and flow rate as parameters at a maximum pressure of 2.6 MPa.
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
Ishida, Shinya; Tagami, Hirotaka; Tobita, Yoshiharu; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
no abstracts in English
Ishii, Katsunori; Morita, Keisuke; Noguchi, Hiroki; Aoki, Takeshi; Mizuta, Naoki; Hasegawa, Takeshi; Nagatsuka, Kentaro; Nomoto, Yasunobu; Shimizu, Atsushi; Iigaki, Kazuhiko; et al.
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2023/09
Okajima, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2023/09
no abstracts in English
Ezure, Toshiki; Akimoto, Yuta; Matsushita, Kentaro; Tanaka, Masaaki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
In hot plenums of sodium-cooled fast reactors, restriction of cover gas entrainment caused by vortex dimples on the free surface is an important thermal-hydraulic issue. For this reason, the authors have developed an evaluation method of gas entrainment with an evaluation tool named, StreamViewer. In this study, evaluation using StreamViewer was applied to a water experiment having a simplified hot pool geometry aiming at the validation of the evaluation method toward the application to the evaluation of a pool-type sodium cooled fast reactor. In StreamViewer, the three-dimensional distribution of pressure decrease along the vortex center line was calculated from the velocity distribution obtained by CFD analyses, and the free surface dimple depth was obtained from the hydraulic balance with the pressure distribution and the cover gas pressure. As the results, it was confirmed that the onset of gas entrainment could be predicted appropriately based on the above-mentioned calculation method.
Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.
Yoshimura, Kazuo; Doda, Norihiro; Nakamine, Yoshiaki*; Fujisaki, Tatsuya*; Igawa, Kenichi*; Iida, Masaki*; Tanaka, Masaaki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
In Japan Atomic Energy Agency, a virtual plant model of the sodium-cooled fast reactor plant composed in a computer is being developed to reduce the development cost, by replacing the experiments to the numerical simulations with coupled analyses of the physical phenomena accounting for the interaction between components under various plant conditions. Through the numerical analysis of the ULOHS test in the U.S. experimental fast reactor named EBR-II, applicability of the virtual plant model was confirmed in comparison with the measured data including the core inlet temperature and the reactor power.
Sogabe, Joji; Kamiyama, Kenji; Tobita, Yoshiharu; Okano, Yasushi
no journal, ,
During severe accidents by an anticipated transient without scram, it is important to evaluate multiphase multi-component flow behavior, when a part of the disrupted core material is discharged outside the disrupted core region through control rod guide tubes. In particular, the blockage behavior of the disrupted core material in a flow path is an important phenomenon that affects the amount of relocated fuels (the fuel discharged outside the disrupted core region and the fuel remaining in the disrupted core region). A fast reactor safety analysis code, SIMMER, is currently being developed for application to the post-accident material relocation (PAMR) phase. In the paper, aiming at actual reactor analyses for the PAMR phase of the SIMMER code, a model for the blockage in the flow path for possible phenomena in the PAMR phase. The model improves the applicability of the SIMMER code to the PAMR phase on the actual reactors.
Tachibana, Yukio; Noguchi, Hiroki; Sumita, Junya; Ohashi, Hirofumi; Sato, Hiroyuki; Sakaba, Nariaki
no journal, ,
For the High-Temperature Gas-cooled Reactor (HTGR) demonstration reactor, establishment of an annular core, enlargement of isolation valves, etc., establishment of codes and standards, establishment of technology to connect the hydrogen production system to the HTGR, etc. are required. In FY2022, JAEA started a project to connect hydrogen production system to the HTTR. JAEA will advance safety evaluations to ensure sufficient safety, and confirm the hydrogen production technology using nuclear heat for the first time in the world. JAEA has been requested by the UK and Poland for technical cooperation for the introduction of HTGRs, and has started technical cooperation to make the best use of the technology for the future demonstration reactor in Japan. This report introduces the current status of development for the HTGR demonstration reactor.
Fujii, Shoma*; Yamano, Hidemasa; Hayafune, Hiroki; Ohno, Shuji
no journal, ,
The application of thermal energy storage (TES) in power generation systems is an attractive option that can convert variable renewable energy, or stable heat sources, such as reactor heat from small modular reactors (SMRs), into flexible power output. A molten salt-based TES charging reactor heat from SMRs was considered as a case study of Power to Heat to Power system, and was feasibly investigated by annual simulation. A price-following model was developed to simulate heat charging and discharging operation, in which reactor heat is stably charged and its heat is discharged during periods of high electricity spot prices. Although the annual output is slightly lower than the SMR stand-alone operation due to the low-load operation of the steam turbine, the total revenue increases due to the improvement of the electricity spot price at the time of power selling.
