Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Mikami, Nao; Aizawa, Kosuke; Ueki, Yoshitaka*; Michel, F.*; Fache, J.*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 10 Pages, 2025/09
Ishida, Shinya; Tagami, Hirotaka*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
Aizawa, Kosuke; Chikazawa, Yoshitaka; Morohoshi, Kyoichi*; Kubo, Koji*; Uchita, Masato*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 8 Pages, 2025/09
Yamano, Hidemasa; Futagami, Satoshi; Sasa, Kyohei*; Nakamura, Hironori*; Tokizaki, Minako*; Kubota, Ryuzaburo*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 12 Pages, 2025/09
This study examined the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on the passive reactor shutdown capability to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Kato, Shinya; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Endo, Tomohiro*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
During a reactor power increase in ULOF and UTOP events in sodium-cooled fast reactors, core deformation due to thermal expansion of core elements is expected to cause a negative feed-back effect to suppress this power increase. An analytical evaluation method of core deformation reactivity for design is being developed in JAEA. However, the neutronics calculation module uses several approximations. This study aims to develop the detailed evaluation method as a reference neutron transport calculation code for confirmation of the validity of the calculated core deformation reactivity. Here, the two-dimensional finite volume method (FVM) code based on simplified P3 (SP3) approximation with unstructured mesh have been developed as the first step of the development. This paper describes the calculation theory of the FVM code, the procedure of introducing SP3 approximation into the code and the verification results of the functions developed.
Onoda, Yuichi; Kubota, Ryuzaburo*; Yamano, Hidemasa
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
Fukuda, Takanari; Uesawa, Shinichiro; Yamashita, Susumu
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 12 Pages, 2025/09
A comparative study was conducted on three interface capturing schemes (ICSs) of the VOF method: THINC/WLIC, THINC/AWLIC, and PLIC for simulating gas-liquid two-phase flow in a BWR reactor core. The simulations in a rod bundle geometry were compared qualitatively and quantitatively with experimental data obtained with a high-speed camera and wire mesh sensors. The results showed that the all ICSs yielded reasonable agreements with experimental data, but THINC/WLIC had a significant issue in which the VOF value diffuses and dissipates over the simulation geometry. THINC/AWLIC, developed by the authors, improved the VOF diffusion issue of the THINC/WLIC and predicted the void fraction close to that of the highly accurate ICS of PLIC, despite its simpler algorithm. However, the numerical bubble coalescence was still an issue, particularly at low gas flow rates, which calls for further research to refine the simulation results to better reflect actual bubble behavior.
Hasegawa, Toshinari; Nagasumi, Satoru; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Nakagawa, Shigeaki; Shimazaki, Yosuke; Nakajima, Kunihiro; Sakurai, Yosuke
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 6 Pages, 2025/09
JAEA has planned a hydrogen production test using the High-Temperature Engineering Test Reactor (HTTR) to demonstrate hydrogen production utilizing the heat from a high-temperature gas-cooled reactor (HTGR). To realize the coupling of a hydrogen production facility with an HTGR, one of the key issues is to confirm the effect of thermal load fluctuations in the facility on the reactor. In this study, a thermal load fluctuation test was conducted during HTTR operation to investigate the reactor's response. The test was performed at 90% reactor power, during which the reactor inlet coolant temperature was increased by 11
C to simulate a thermal load fluctuation. As a result, the reactor outlet coolant temperature remained almost unchanged, and the heat corresponding to the inlet temperature increase was absorbed by the core graphite blocks. Furthermore, due to the negative reactivity feedback effect associated with the rise in graphite block temperature, the reactor power decreased to 88% and stabilized without any control rod operation. These findings indicate that disturbances in the reactor inlet coolant temperature are mitigated by the heat storage capacity of the core graphite blocks.
Imaizumi, Yuya; Sogabe, Joji*; Onoda, Yuichi; Yamano, Hidemasa
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 9 Pages, 2025/09
Ono, Ayako; Okawa, Tomio*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 12 Pages, 2025/09
We aim to establish a new prediction method for departure from nucleate boiling (DNB) in fuel assemblies, contributing to the design and safety evaluation of next-generation reactors. Since the formation of a large vapor mass on the heating surface is considered the initial trigger of DNB, we focus on predicting the heat flux required for its formation. To achieve this, we integrate two-phase flow computational fluid dynamics (CFD) with a mechanistic large vapor mass formation model. In this study, we develop a simplified numerical analysis method that simulates bubble generation, growth, motion, and coalescence to predict large vapor mass formation. Additionally, we investigate the vapor volume on the heating surface at the onset of large vapor mass formation and predict their formation conditions.
Hasegawa, Toshinari; Nagasumi, Satoru; Nemoto, Takahiro; Nakajima, Kunihiro; Yokoyama, Keisuke; Fujiwara, Yusuke; Arakawa, Ryoki; Iigaki, Kazuhiko; Inoi, Hiroyuki; Kawamoto, Taiki
Proceedings of 2024 International Congress on Advanced in Nuclear Power Plants (ICAPP 2024) (Internet), 10 Pages, 2024/06
The filter element of the primary gas circulators (PGC) in High Temperature engineering Test Reactor (HTTR) and its deposits were investigated by Scanning Electron Microscope (SEM) observation and Energy Dispersive X-ray spectroscopy (EDX) analysis to find the cause of the increase of the filter differential pressure during the operation in 2021. SEM observation showed that the clumpy deposits and fibrous deposits smaller than the filtration pore size and the rod-shaped deposits larger than the pore size were present on the filter element. EDX analysis showed that the clumpy deposits and fibrous deposits could include silicone oil in the primary helium purification system (PHPS) gas circulators and that the rod-shaped deposits were thermal insulators inside of the co-axial double pipes in the primary cooling system. It is considered that silicone oil leaked from the PHPS gas circulators due to deterioration in the absorption performance of the activated charcoal filter. Next, it could be vaporized and reach PGC's filter element after passing through the reactor core. Since those deposits including silicone oil were present over the entire surface of the filter element, the filter differential pressure could be increased due to a reduction in the pore size and a rise in its flow resistance. The thermal insulator was unrelated to filter clogging because it was present mainly in the lower part of the filter element. Therefore, silicone oil could increase the filter differential pressure, and the graphite powder, which is the cause of the previous issue was unrelated.
Yamamoto, Tomohiko; Kato, Atsushi; Hayakawa, Masato; Shimoyama, Kazuhito; Ara, Kuniaki; Hatakeyama, Nozomu*; Yamauchi, Kanau*; Eda, Yuhei*; Yui, Masahiro*
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 6 Pages, 2023/04
Takano, Kazuya; Oki, Shigeo; Doda, Norihiro; Chikazawa, Yoshitaka; Maeda, Seiichiro
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 7 Pages, 2023/04
The MOX fueled SMR-SFRs with lower linear heat rating of 100 W/cm and 50 W/cm, whereas the linear heat rating at rated power is around 400 W/cm in general, were designed to decrease the fuel temperature during its rated power state in order to pursue the inherent core safety for MOX fueled SMR-SFRs. The transient analyses for Anticipated Transient Without Scram (ATWS) events represented by an Unprotected Loss of Flow (ULOF) accident on the lower linear heat rating cores were performed considering their inherent feedback reactivity. Through the transient analysis, the inherent core safety performances for the lower linear heat rating cores were discussed based on the evaluated maximum coolant temperature and Cumulative Damage Fraction (CDF) as criteria to maintain the core and fuel integrity. The feasible design window for MOX fueled SMR-SFRs with the inherent core safety focusing on the linear heat rating was identified based on the transient analysis results.
Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
A series of Fuel-Coolant Interaction (FCI) experiments discharging about 1 kg of molten stainless steel (SS), one of prototypic materials, into a sodium pool (about 30 cm in diameter, about 1 m in depth) has been conducted at the MELT facility of Japan Atomic Energy Agency. X-ray visualization program confirmed that molten SS jets with diameters of about 1 to 2 cm were broken up by FCIs which occurred with sodium vapor expansion at depths near the initial sodium level. Responses of thermocouples suggested that the jets were rapidly cooled near the initial sodium level and became SS fragments before reaching the bottom of the sodium pool. Results of experiments with molten SS jets confirmed that an evaluation methodology for jet breakup under sodium-cooled conditions needs to be developed by taking into account effects of thermal parameters related to sodium vaporization as well as conventional parameters, such as density ratio of a jet and coolant, and jet diameter.
Tagami, Hirotaka; Tobita, Yoshiharu
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
Recently, the safety analysis for a licensing of small nuclear power fast reactor is performed. It is necessary to confirm the effectiveness of the design measure to prevent the CV failure in the licensing procedure. Because the energy generation in TP of ULOF is one of the main factors to affect the integrity of CV, the ULOF behavior is analyzed using SIMMER developed under international cooperation. Although the characteristic of TP in small reactor is a slow and mild event progression due to the negative void reactivity, several conservative assumptions are applied in the analysis. Because the prompt criticality by fuel compaction is mainly driven by a fuel coolant interaction, its impact on energy generation is also investigated by conservatively assuming uncertainties. The obtained results by the analysis using SIMMER are used for the subsequent phase to analyze the mechanical integrities of reactor vessel and CV.
Koike, Akari*; Nemoto, Masaya*; Nakashima, Risako*; Sakai, Takaaki*; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 2 Pages, 2023/04
To evaluate the effect of the operator's recognition of the accident management (AM) necessity on plant safety, the operator's recognition of the AM necessity was modeled as a function of time-dependent success probability, and dynamic PRA analyses were performed for a sodium-cooled fast reactor during abnormal snowfall. The analysis results showed that the operator's recognition of the snowfall can avoid the core damage at an earlier stage after the accident.
Takamatsu, Kuniyoshi; Funatani, Shumpei*
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 17 Pages, 2023/04
The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Therefore, the authors concluded that the proposed RCCS based on atmospheric radiation has the advantage that the temperature of the RPV can be stably maintained against disturbances in the outside air (ambient air). Moreover, methodology to utilize all the heat emitted from the RPV surface for increasing the degree of waste-heat utilization was discussed.
Nishida, Akemi; Kawata, Manabu; Choi, B.; Iigaki, Kazuhiko; Shiomi, Tadahiko; Li, Y.
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 10 Pages, 2023/04
Our research and development are aimed at improving the accuracy of the three-dimensional seismic evaluation analysis method for nuclear buildings, which contributes to the probabilistic risk assessment caused by earthquakes (seismic PRA), using actual seismic observation records in collaboration with the Nuclear Regulation Authority since 2019. In this study, we constructed a large-scale observation system that enabled simultaneous measurements at multiple positions during earthquakes or against artificial waves. The accelerometers of the observation system were installed on/in the soil, floors, and on the walls of the nuclear building. This paper presents an outline of the large-scale observation system and the findings obtained from the analysis of the seismic observation records acquired using the system.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
Higuchi, Kyoko*; Kurita, Keisuke; Sakai, Takuro; Suzui, Nobuo*; Sasaki, Minori*; Katori, Maya*; Wakabayashi, Yuna*; Majima, Yuta*; Saito, Akihiro*; Oyama, Takuji*; et al.
Plants (Internet), 11(6), p.817_1 - 817_11, 2022/03
Times Cited Count：3 Percentile：16.22(Plant Sciences)Genetic diversity in the rate of Fe uptake by plants has not been broadly surveyed among plant species or genotypes, although plants have developed various Fe acquisition mechanisms. We adopted the "Live-autoradiography" technique with radioactive 
Fe to directly evaluate the uptake rate of Fe by barley cultivars from a nutrient solution containing a very low concentration of Fe. Our observations revealed that the ability to acquire Fe from the low Fe solution was not always the sole determinant of tolerance to Fe deficiency among the barley genotypes.
