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Rizaal, M.; Nakajima, Kunihisa; Suzuki, Eriko; Miwa, Shuhei
Annals of Nuclear Energy, 218, p.111433_1 - 111433_10, 2025/08
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：0JAEA 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
Hasegawa, Kunio; Yamaguchi, Yoshihito; Udyawar, A.*
Journal of Pressure Vessel Technology, 147(3), p.034501_1 - 034501_7, 2025/06
Satou, Akira; Wada, Yuki; Shibamoto, Yasuteru
Nuclear Engineering and Design, 437, p.114020_1 - 114020_14, 2025/06
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.
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：0.00(Thermodynamics)
for negative stress ratios R based on trend in experimental data for fatigue crack growth rates of austenitic stainless steels for ASME code Section XI Negyesi, 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：0Futemma, 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.
Terada, Atsuhiko; Thwe Thwe, A.; Hino, Ryutaro*
JAEA-Review 2024-049, 400 Pages, 2025/03
JAEA-Review-2024-049.pdf:13.94MB
In the aftermath of the Fukushima Daiichi Nuclear Power Station accident, safety measures against hydrogen in severe accident has been recognized as a serious technical problem in Japan. As one of efforts to form a common knowledge base between nuclear engineers and experts on combustion and explosion, we issued the "Handbook of Advanced Nuclear Hydrogen Safety (1st edition)" in 2017. For improvement of the rational advancement of hydrogen safety measures and further reliability of hydrogen safety evaluation, a CFD analysis is highly expected to produce more precisely and quantitative results. We have been developing an integrated CFD analysis code system which can analyze hydrogen diffusion, explosion-combustion and structural integrity at the severe accident especially for pressurized water reactors (PWRs). We organized the role of LP and the CFD analyses and their utilization examples of hydrogen safety validation. Based on these results, we made the "Handbook of Advanced Nuclear Hydrogen Safety (2nd volume)". The analysis results of real scale PWR described in 2nd volume are confirmed by cross-analysis models and existing data obtained through representative small, medium and large-scale tests.
Hirouchi, Jun; Watanabe, Masatoshi*; Hayashi, Naho; Nagakubo, Azusa; Takahara, Shogo
JAEA-Research 2024-015, 114 Pages, 2025/03
JAEA-Research-2024-015.pdf:10.03MB
The public living in areas contaminated by nuclear accidents is exposed to radiation in the early phase and over the long term. Even under the same accident scenario, the exposure doses and the effectiveness of sheltering, which is one of the protective measures, vary depending on the meteorological condition and the surrounding environment. The exposure doses and sheltering effectiveness in the early phase are important information for the public and the national and local governments planning a nuclear emergency preparedness. In this report, we evaluate the exposure doses and sheltering effectiveness at sites with nuclear facilities in Japan using OSCAAR, one of the probabilistic risk assessment codes, for five accident scenarios: three scenarios from past severe accident studies; a scenario defined by the Nuclear Regulatory Authority; and a scenario assuming the Fukushima Daiichi Nuclear Power Station accident. The sheltering effectiveness differed by approximately 20% among the sites. This was due to the differences in wind speed among the sites.
Inoue, Yuki; Yamada, Junya; Hamaguchi, Takumi; Seya, Natsumi; Muto, Yasunobu; Nohara, Naofumi; Oishi, Tetsuya; Hashimoto, Makoto
JAEA-Data/Code 2024-017, 109 Pages, 2025/03
JAEA-Data-Code-2024-017.pdf:4.54MB
This report presents the results of environmental radiation monitoring at Oarai Nuclear Engineering Institute, JAEA for 10 years from 2011 to 2021 after the accident at the Fukushima Daiichi Nuclear Power Station. Database of monitoring results are included as an appendix. Environmental transfer parameters, deposition velocity, resuspension factor, concentration factor and distribution coefficient, are calculated as an Appendix.
Scaria, J.*; P
drot, M.*; Fablet, L.*; Yomogida, Takumi; Nguyen, T. T.*; Sivry, Y.*; Catrouillet, C.*; Pradas del Real, A. E.*; Choueikani, F.*; Vantelon, D.*; et al.
Environmental Science & Technology, 59(11), p.5747 - 5755, 2025/03
Times Cited Count：0Understanding and predicting the interaction mechanisms between chromium and magnetite is of particular interest to elucidate the biogeochemical behavior of Cr in the environment and to develop optimal soil remediation and water treatment strategies. However, while the elimination of the most toxic form of (Cr(VI)) by its reduction to Cr(III) has widely been documented, elucidating the exact mechanism involved in Cr(III) sorption to magnetite has attracted less attention. This study examined the interaction of Cr(III) solution with 10 nm-sized magnetites, whose stoichiometries were carefully defined and preserved in anaerobic conditions. This study reveals the joint effects of pH and magnetite stoichiometry on Cr(III) sorption mechanism, and that Cr(III)-(hydr)oxide precipitation is not necessarily the driving process of Cr(III) elimination from solutions. These results will help predict the fate and transport of chromium, as well as developing magnetite-based chromium remediation processes.
Hirouchi, Jun; Kujiraoka, Ikuo; Takahara, Shogo; Takada, Momo*; Schneider, T.*; Kai, Michiaki*
Journal of Radiological Protection, 45(1), p.011508_1 - 011508_14, 2025/03
Times Cited Count：0no abstracts in English
Hirouchi, Jun; Watanabe, Masatoshi*; Hayashi, Naho; Nagakubo, Azusa; Takahara, Shogo
Journal of Radiological Protection, 45(1), p.011506_1 - 011506_11, 2025/03
Times Cited Count：0Public living in areas contaminated by nuclear accidents is exposed to radiation in the early phase and over the long term. Even under similar accident scenarios, radiation doses and sheltering effectiveness, which is one of the protective measures, depend on meteorological conditions and the surrounding environment. Radiation doses and sheltering effectiveness in the early phase of nuclear accidents are crucial information for the public as well as national and local governments planning a nuclear emergency preparedness. In this study, we assessed radiation doses and sheltering effectiveness at sites with nuclear facilities in Japan using the Off-Site Consequence Analysis code for Atmospheric Release accidents, which is one of the level-3 probabilistic risk assessment codes, for five accident scenarios: three scenarios from past severe accident studies, a scenario defined by the Nuclear Regulation Authority in Japan, and a scenario corresponding to the Fukushima-Daiichi Nuclear Power Station accident. The sheltering effectiveness differed by up to approximately 50% among the accident scenarios at the same sites and by approximately 20%
50% among sites under the same accident scenario. Differences in the radionuclide composition among the accident scenarios and the differences in wind speeds among the sites primarily caused these differences in sheltering effectiveness.
Watanabe, Kaho; Suzuki, Kenta
JAEA-Testing 2024-004, 41 Pages, 2025/02
JAEA-Testing-2024-004.pdf:3.4MB
There is an emergency response team against nuclear facilities' accidents of Japan Atomic Energy Agency (JAEA). The team is in Naraha Center for Remote Control Technology Development (NARREC). Trainings of maneuvering emergency response robots are performed at NARREC by the team for operators temporarily coming from all facilities which are possibility to have a huge nuclear emergency in JAEA. There are three kinds of robots to be used in the training: (1) Reconnaissance robots (for shooting a video and for measuring radiation level and temperature in nuclear accidents field), (2) Work robots (for obstacles removal in nuclear accidents field) and (3) Drones. As for the training, each operator person has been trained once or twice a year. Therefore, it is difficult for operators to retain their own sense of the maneuvering robots. Then, the team provided a simulator system which is consists of a palm-sized controller and a PC installed operation programs based on Choreonoid as dynamics simulation software, and we provided the system for operators in order to supplement a real robot operation training. This report shows the operation method of the provided simulator system regarding one reconnaissance robot and two work robots.
Okagaki, Yuria; Hibiki, Takashi*; Shibamoto, Yasuteru
JAEA-Review 2024-047, 58 Pages, 2025/02
JAEA-Review-2024-047.pdf:2.22MB
In PWR accident scenarios, the injection of water from the ECCS (ECC injection) might result in thermal stratification in the case of the insufficient mixing of cold and hot water and induce a PTS, affecting the RPV integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal-hydraulic and structural analysis. Especially in the thermal-hydraulic approach, reliable CFD simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the DC. This study reviewed from the viewpoint of the turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD simulation for the experiment on PTS performed in the ROCOM, Transient TOPFLOW, UPTF, and LSTF.
Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 237, p.126406_1 - 126406_15, 2025/02
Times Cited Count：1 Percentile：0.00(Thermodynamics)Sun, Haomin; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 237, p.126445_1 - 126445_14, 2025/02
Times Cited Count：0 Percentile：0.00(Thermodynamics)Fujita, Tatsuya; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 62(2), p.179 - 196, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study newly established a direct coupling code system consisting of the nuclear data processing code FRENDY version 2, and the three-dimensional heterogeneous transport code GENESIS (FRENDY-V2/GENESIS) for easy implementation of the random-sampling-based uncertainty quantification considering the implicit effect due to nuclear cross-section (XS) perturbations. The multi-group macroscopic XSs prepared for GENESIS were generated by FRENDY version 2, where the Dancoff factor was calculated by the neutron current method. Then the background XSs were evaluated based on the Carlvik two-term rational approximation. The infinite multiplication factor (k-infinity) and the fission reaction rate distribution in UO
and MOX lattice geometries were compared with MVP3 to verify the calculation accuracy of FRENDY-V2/GENESIS. The sensitivity analyses on the discretization conditions such as the ray tracing of the method of characteristics were also carried out. Through several comparisons between FRENDY-V2/GENESIS and MVP3, FRENDY-V2/GENESIS with the SHEM 361-group structure calculates the k-infinity within approximately 50 pcm and the fission reaction rate distribution within approximately 0.1% by the root mean square, respectively. Consequently, the applicability of FRENDY-V2/GENESIS was verified, and FRENDY-V2/GENESIS can be used to discuss the implicit effect due to multi-group XS perturbations.
