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Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 249, p.127219_1 - 127219_16, 2025/10
Soma, Shu; Ishigaki, Masahiro*; Shibamoto, Yasuteru
Annals of Nuclear Energy, 219, p.111455_1 - 111455_12, 2025/09
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Wang, Z.; Matsumoto, Toshinori; Shibamoto, Yasuteru; Duan, G.*
Journal of Computational Physics, 537, p.114072_1 - 114072_29, 2025/09
Rizaal, M.; Nakajima, Kunihisa; Suzuki, Eriko; Miwa, Shuhei
Annals of Nuclear Energy, 218, p.111433_1 - 111433_10, 2025/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Araki, Shohei; Aizawa, Eiju; Murakami, Takahiko; Arakaki, Yu; Tada, Yuta; Kamikawa, Yutaka; Hasegawa, Kenta; Yoshikawa, Tomoki; Sumiya, Masato; Seki, Masakazu; et al.
Annals of Nuclear Energy, 217, p.111323_1 - 111323_8, 2025/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)JAEA has modified the STACY from a homogeneous system using solution fuel to a heterogeneous system using fuel rods in order to obtain criticality characteristics of fuel debris. The modification of the STACY was completed in December 2023. A series of performance inspections were conducted for the start of experimental operations. A new thermal power calibration is required for the performance inspections in order to operate at less than 200 W, which is the permitted thermal power. However, the thermal power measurement method and calibration data used in the former STACY is no longer available due to the modification of the modified STACY. We measured the thermal power of the STACY using the activation method that was improved to adapt to the measurement condition and calibrated the power meter system. Since the positions where activation foils could be installed were very limited, the thermal power was evaluated using numerical calculations supplemented by experimental data. Neutron flux data at the positions of the activation foil was measured by the activation method. Neutron distribution in the core was calculated by the Monte Carlo code MVP. A response function of the activation foil was calculated using the PHITS. The uncertainty of the thermal power measurement was conservatively estimated to be about 15%. Four operations were conducted for the thermal power measurement. The power meter was calibrated by using three operational data and tested with the one operational data. It was found that the indicated value of the meter adjusted by the STACY before the modification work would tend to overestimate the actual output by about 40%. In addition, the current calibration was able to calibrate the meter to within 3% accuracy.
Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru
Progress in Nuclear Energy, 185, p.105771_1 - 105771_13, 2025/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Yoshida, Kazuo; Hiyama, Mina*; Tamaki, Hitoshi
JAEA-Research 2025-003, 24 Pages, 2025/06
JAEA-Research-2025-003.pdf:2.06MB
An accident of evaporation to dryness by boiling of high-level radioactive liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (RuO
) are released from the tanks with water and nitric-acid mixed vapor into the atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. RuO is expected to be absorbed chemically into water dissolving nitrous acid. Condensation of mixed vapor plays an important role for Ru transporting behavior in the facility building. The thermal-hydraulic behavior in the facility building is simulated with MELCOR code. The latent heat, which is a governing factor for vapor condensing behavior, has almost same value for nitric acid and water at the temperature range under 120 centigrade. Considering this thermal characteristic, it is assumed that the amount of nitric acid is substituted with mole-equivalent water in MELCOR simulation. Compensating modeling induced deviation by this assumption have been assembled with control function features of MELCOR. The comparison results have been described conducted between original simulation and modified simulation with compensating model in this report. It has been revealed that the total amount of pool water in the facility was as same as both simulations.
Iwasawa, Yuzuru; Matsumoto, Toshinori; Moriyama, Kiyofumi*
JAEA-Data/Code 2025-001, 199 Pages, 2025/06
JAEA-Data-Code-2025-001.pdf:9.11MB
A steam explosion is defined as a phenomenon that occurs when a hot liquid comes into contact with a low-temperature cold liquid with volatile properties. The rapid transfer of heat from the hot liquid to the cold liquid results in a chain reaction of the explosive vaporization of the cold liquid and fine fragmentation of the hot liquid. The explosive vaporization of the cold liquid initiates the propagation of shock waves in the cold liquid. The expansion of the hot and cold liquid mixture exerts mechanical forces on the surrounding structures. In severe accidents of light water reactors, the molten core (melt) is displaced into the coolant water, resulting in fuel-coolant interactions (FCIs). The explosive FCI, referred to as a steam explosion, has been identified as a significant safety assessment issue as it can compromise the integrity of the primary containment vessel. The JASMINE code is an analytical tool developed to evaluate the mechanical forces imposed by steam explosions in nuclear reactors. It performs numerical simulations of steam explosions in a mechanistic manner. The present report describes modeling concepts, basic equations, numerical solutions, and example simulations, as well as instructions for input preparation, code execution, and the use of supporting tools for practical purpose. The present report is the updated version of the "Steam Explosion Simulation Code JASMINE v.3 User's Guide, JAEA-Data/ Code 2008-014". The present report was compiled and updated based on the latest version of the code, JASMINE 3.3c, with corrections for minor errors of the distributed code JASMINE 3.3b, and conformance to recently widely used compilers on UNIX-like environments such as the GNU compiler. The numerical simulations described in the present report were obtained using the latest version JASMINE 3.3c. The latest parameter adjustment method for a model parameter proposed by the previous study is employed to conduct the numerical simulations.
Hasegawa, Kunio; Yamaguchi, Yoshihito; Udyawar, A.*
Journal of Pressure Vessel Technology, 147(3), p.034501_1 - 034501_7, 2025/06
Times Cited Count：0Satou, Akira; Wada, Yuki; Shibamoto, Yasuteru
Nuclear Engineering and Design, 437, p.114020_1 - 114020_14, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Post-boiling transition (post-BT) heat transfer is essential for analyzing the duration of surface dryout and peak cladding temperature during abnormal transients and accidents in light water reactors. The rewetting phenomenon is very important for evaluating the dryout duration. However, due to the lack of an experimental database on rewetting velocities under high flow and heat flux conditions, sufficient data for model development and validation do not exist. Therefore, a database on rewetting velocities caused by stepwise boundary condition changes under a wide range and multiple combination of thermal-hydraulic conditions was obtained using a single-tube experimental apparatus. Based on this database and the characteristics of rewetting velocities obtained, an experimental correlation for rewetting velocity was proposed. This correlation predicts the rewetting velocity accurately by taking the change in the mass flux of the liquid or gas phase with stepwise transients as a parameter. This suggested that the change in the mass flux of the gas or liquid phase near the liquid film front has a strong influence on the rewetting under extremely high mass flux conditions compared to the reflooding process.
Mori, Airi; Johansen, M. P.*; McGinnity, P.*; Takahara, Shogo
Communications Earth & Environment (Internet), 6, p.356_1 - 356_11, 2025/05
Times Cited Count：0Fablet, L.*; P
drot, M.*; Choueikani, F.*; Kieffer, I.*; Proux, O.*; Pierson-Wickmann, A.-C.*; Cagniart, V.*; Yomogida, Takumi; Marsac, R.*
Environmental Science; Nano, 12(5), p.2815 - 2827, 2025/05
Times Cited Count：0 Percentile：0.00(Chemistry, Multidisciplinary)Nickel is an omnipresent trace element in the environment. Due to its high affinity for iron oxide nanoparticles, its elimination from soils and water by these nanoparticles represents an interesting strategy, specially by magnetites, which is naturally present in the environment. However, the interactions between Ni and magnetite are poorly understood, because of the difficulty to control the stoichiometry (Fe(II)-to-Fe(III) ratio) of magnetite. The behavior of Ni in the presence of magnetite nanoparticles with different stoichiometries, in aqueous solution and inert atmosphere, are probed by adsorption experiments and X-ray Absorption Spectroscopy. This study helps predicting the interactions between Ni and magnetite in environmental conditions, which can be used for the development of efficient remediation strategies.
Hirouchi, Jun; Takahara, Shogo; Watanabe, Masatoshi*
Journal of Radiological Protection, 45(2), p.021506_1 - 021506_13, 2025/05
Times Cited Count：0Sheltering is a key countermeasure for mitigating radiation exposures during nuclear power plant accidents. The effectiveness of sheltering in minimizing inhalation exposure is commonly expressed using the reduction factor, which is the ratio of indoor to outdoor cumulative doses. The indoor dose is primarily influenced by the air exchange rate, penetration factor, and indoor deposition rate. Additionally, the air exchange rate is dependent on wind speed. In previous studies, the reduction factor was often treated as a constant value or calculated under constant wind speed conditions. However, wind speed varies in reality. This study investigated the effect of temporal variations in wind speed on the reduction factor and developed a simplified correction method to account for these variations. The results revealed that temporal variations in wind speed caused the reduction factor to differ by a factor of approximately two. Using the simplified correction method, the corrected reduction factors agreed, on average, within 10% of those calculated using a method that explicitly considers temporal variations in actual wind speed. Additionally, the computational cost was reduced by more than 20 times.
Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 239, p.126598_1 - 126598_18, 2025/04
Times Cited Count：1 Percentile：27.01(Thermodynamics)Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 16 Pages, 2025/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study investigated the impact of nuclear data updates from JENDL-4.0 (J4) to JENDL-5 (J5) on the light-water reactor fuel burnup calculations. Burnup calculations were conducted with J4 and J5 for PWR pin-cell and BWR fuel assembly geometries. The calculation results revealed significant burnup-dependent differences in the neutron multiplication factor (k
). Across the burnup range of 0-50 GWd/t, k
values of J5 were consistently smaller than those of J4 and the difference gradually increased as burnup progressed. Direct sensitivity calculations, in which each nuclide data was replaced from J4 to J5, indicated that updates to the cross-sections of 
U, 
U, and 
Pu and the thermal scattering law data of H in H
O notably impacted the k differences. For the BWR assembly geometry containing Gd fuels, large k differences were observed in the burnup range of 10-15 GWd/t. This difference was primarily attributed to updates in the U, 
Gd, and 
Gd cross-sections, and thermal scattering law data of H in HO. Furthermore, we investigated how the nuclear data updates affected the k differences by examining nuclide number densities, the energy-dependent sensitivities, and the neutron spectra.
for negative stress ratios R based on trend in experimental data for fatigue crack growth rates of austenitic stainless steels for ASME code Section XINegyesi, M.*; Yamaguchi, Yoshihito; Hasegawa, Kunio; Lacroix, V.*; Morley, A.*
Journal of Pressure Vessel Technology, 147(2), p.021201_1 - 021201_7, 2025/04
Times Cited Count：0 Percentile：0.00(Engineering, Mechanical)Shiotani, Kohei; Niiyama, Tomoaki*; Shimokawa, Tomotsugu*
Materials Transactions, 66(6), p.704 - 711, 2025/04
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)no abstracts in English
Watanabe, Tomoaki; Aizawa, Naoto*; Chiba, Go*; Tada, Kenichi; Yamamoto, Akio*
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025) (Internet), 10 Pages, 2025/04
Currently, a major burnup calculation method for the nuclide composition of nuclear fuel conducts neutron transport calculations at each burnup step to account for changes in the neutron spectrum. While this method is highly accurate, the large computational cost of neutron transport calculations can be problematic. Therefore, a fast burnup calculation method based on neutron spectrum reconstruction with the proper orthogonal decomposition (POD) and regression model is investigated. In this method, dimensionality reduction by POD is applied to many neutron fluxes obtained from detailed burnup calculations for various input parameter sets, and regression models are constructed to connect the dimensionality-reduced neutron fluxes and parameters. By substituting arbitrary input parameters to the regression models, the neutron flux is reconstructed and the burnup calculation is performed. This method performs burnup calculations that consider changes in the neutron spectrum based on input conditions without neutron transport calculations. The present method was applied to a PWR UO fuel pin cell model. The results show the nuclide inventory can be calculated with a prediction accuracy within a few percent. In addition, it is found that the calculation error is dominated by the regression models, which implies the further improvement of the regression models leads to improving the accuracy.
Futemma, Akira; Sanada, Yukihisa; Nakama, Shigeo; Sasaki, Miyuki; Ochi, Kotaro; Sawahata, Yoshiro*; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; Haginoya, Masashi*; et al.
JAEA-Technology 2024-022, 170 Pages, 2025/03
JAEA-Technology-2024-022.pdf:15.09MB
On March 11, 2011, the 2011 off the Pacific coast of Tohoku Earthquake caused a tsunami that led to the Fukushima Daiichi Nuclear Power Station accident, releasing radioactive material into the environment. Since then, Aerial Radiation Monitoring (ARM) using manned helicopters has been employed to measure radiation distribution. As a commissioned project from the Nuclear Regulation Authority, the Japan Atomic Energy Agency (JAEA) utilizes this technology for emergency monitoring during nuclear facility accidents, aiming to provide prompt results by pre-arranging information on background radiation, topography, and control airspaces around nuclear power plants nationwide. In fiscal year 2023, the commissioned project included conducting ARM around the Sendai Nuclear Power Station and preparing related information. To enhance effectiveness during emergencies, ARM and the first domestic training flight of Unmanned Aerial Vehicles (UAVs) were conducted during the FY2023 Nuclear Energy Disaster Prevention Drill. Furthermore, UAVs radiation monitoring technology was advanced by selecting UAVs and investigating their performance. This report summarizes the results and technical issues identified providing insights to improve emergency preparedness.
Futemma, Akira; Sanada, Yukihisa; Nakama, Shigeo; Sasaki, Miyuki; Ochi, Kotaro; Nagakubo, Azusa; Sawahata, Yoshiro*; Kawasaki, Yoshiharu*; Iwai, Takeyuki*; Hiraga, Shogo*; et al.
JAEA-Technology 2024-021, 232 Pages, 2025/03
JAEA-Technology-2024-021.pdf:25.79MB
The 2011 off the Pacific coast of Tohoku Earthquake on March 11, 2011, caused a tsunami that led to the TEPCO's Fukushima Daiichi Nuclear Power Station (FDNPS) accident, releasing a large amount of radioactive material into the surrounding environment. Since the accident, Aerial Radiation Monitoring (ARM) has been used to quickly and widely measure radiation distribution. As a commissioned project from the Nuclear Regulation Authority, the Japan Atomic Energy Agency (JAEA) has continuously conducted ARM around FDNPS using manned and unmanned helicopters. This report summarizes the monitoring results for fiscal year 2023, evaluates changes in dose rate from past results, and discusses the factors contributing to these changes. Additionally, an analysis considering terrain undulation was conducted to improve accuracy for converting ARM data into dose rate. Furthermore, a method to discriminate airborne radon progeny was applied for ARM results to evaluate its impact. Moreover, to perform wide-area monitoring more efficiently, we advanced the development of unmanned airplane monitoring technology.
