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Sato, Takuto; Nakayama, Hiromasa; Satoh, Daiki
Journal of Nuclear Science and Technology, 63(4), p.426 - 442, 2026/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)We developed a framework for rapid monitoring of radioactive plumes in the vicinity of nuclear facilities based on a quick and practical high-resolution atmospheric dispersion simulation method that combines a large-eddy simulation (LES) model pre-simulation database (pre-sim DB) of wind conditions and onsite meteorological observation results, as proposed by the previous study. However, this framework was not quantitatively demonstrated using measurement data. In this study, we evaluated the performance of the wind condition reproduction and plume dispersion analysis methods. Air dose rates observed at monitoring posts around the stack were compared with the values reproduced by the method using the pre-sim DB, and the reproducibility of both air dose rate and flow field was discussed. The pre-sim DB-based method successfully captured the temporal variation of air dose rates at the monitoring posts, though it tended to overestimate the peak values. Particularly when the vertical wind shear was pronounced, the method using the pre-sim DB could cause significant errors. This is likely because the method relies on wind conditions from a single observation point, which inherently limits its ability to represent vertical wind shear within the pre-sim DB. Despite these limitations, particularly in reproducing complex wind fields, the method utilizing the pre-sim DB offers a valuable and practical tool for rapid dose rate simulation due to its lower computational cost compared to unsteady simulations using an LES model.
Tomita, Jumpei; Tomita, Ryohei; Suzuki, Daisuke; Yasuda, Kenichiro; Miyamoto, Yutaka
Journal of Nuclear Science and Technology, 63(4), p.443 - 454, 2026/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Manabe, Kentaro; Murota, Shuhei; Takahashi, Fumiaki
Journal of Nuclear Science and Technology, 63(4), p.477 - 486, 2026/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)We have developed an internal dose assessment code (IDCC) based on the latest dose assessment methods in accordance with the ICRP 2007 Recommendations. This code enables calculation of effective dose coefficients and intake estimation from individual monitoring results for all radioactive nuclides including short-lived nuclides. The validity of the code has been confirmed through comparisons with the dose coefficient database published by the ICRP and multiple literature examples. We plan to enhance the code for evaluating public exposure, and it is expected to contribute to the revision of regulatory standards for radiation protection and serve as a practical dose assessment tool based on new standards.
I neutron capture cross-section in the keV neutron regionRovira Leveroni, G.; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Iwamoto, Osamu; Iwamoto, Nobuyuki; Katabuchi, Tatsuya*
Journal of Nuclear Science and Technology, 63(4), p.358 - 369, 2026/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Toigawa, Tomohiro; Hotoku, Shinobu; Kumagai, Yuta; Abe, Yuma*; Oyama, Kanichi*; Fukaya, Hiroyuki; Ban, Yasutoshi; Kida, Takashi; Hasegawa, Satoshi*; Nakano, Masanao*; et al.
Journal of Nuclear Science and Technology, 63(3), p.322 - 327, 2026/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The effect of temperature on hydrogen production generated from radiolysis was investigated to determine the associated implications for nuclear fuel reprocessing safety. The hydrogen yield from radiolysis of plutonium nitric acid solution was measured at temperatures up to the boiling temperature of the solution. The results showed no notable temperature dependence even under boiling conditions. The impact of solution agitation on hydrogen production was also assessed, which revealed minor differences in the hydrogen yield between static and agitated conditions at room temperature. These findings suggest that high temperatures or boiling the solution do not considerably enhance hydrogen generation, and provide crucial information for accurately modeling hydrogen risks under severe accidents.
Watanabe, Tomoaki; Aizawa, Naoto*; Chiba, Go*; Tada, Kenichi; Fujita, Tatsuya*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 63(2), p.166 - 186, 2026/02
Times Cited Count：1 Percentile：0.00(Nuclear Science & Technology)A fast burnup calculation method based on neutron spectrum reconstruction is proposed. The method employs a reduced-order model (ROM), constructed using proper orthogonal decomposition (POD) and regression models, to estimate neutron spectra experienced by fuel during burnup. The ROM is built from snapshot data generated through detailed burnup and neutron transport simulations under various conditions. During burnup calculations, the ROM is used to rapidly reconstruct neutron spectra at each burnup step. These reconstructed spectra are then used to compute one-group cross sections from multi-group effective cross sections derived using background cross sections. The proposed method significantly reduces computational time by avoiding repeated neutron transport simulations. Its performance is demonstrated using a PWR UO
fuel pin model. Results show that, with the 6th-order POD, the method predicts nuclide inventories with an average error within 
5% compared to reference Monte Carlo calculations. Error analysis indicates that prediction accuracy is primarily limited by the regression models, rather than by the POD truncation or the multi-group cross section calculations.
Gunji, Satoshi; Araki, Shohei
Journal of Nuclear Science and Technology, 63(2), p.187 - 199, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)
Th as a long-life 
Ac generator using the experimental fast reactor JoyoSasaki, Yuto; Maeda, Shigetaka; Fukasawa, Tetsuo*; Takaki, Naoyuki*
Journal of Nuclear Science and Technology, 63(2), p.154 - 165, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In recent years, targeted alpha therapy, which utilizes Ac combined with antibodies or peptides that selectively accumulate in cancer cells, has garnered attention in the field of nuclear medicine. To meet the resulting increasing demand for Ac, exploring alternative production methods is essential. While several researchers, including the authors, have explored production methods using 
Ra as a raw material, challenges remain, such as the limited availability of Ra, difficulties in handling it, and the requirement for regular irradiation. To address these challenges, the authors focused on developing a production strategy for a long-life Ac generator using 
Th as a raw material and the experimental fast reactor Joyo. A detailed investigation was conducted, encompassing chemical processing after irradiation, target availability, and production yields, including the most probable values and associated uncertainties. Results revealed that although enrichment of the raw material and long-term irradiation are required, Ac can be produced in quantities comparable to its current global supply. Furthermore, this research has shown that the THOREX method, which is already in practical use, be applied to effectively separate by-products such as fission products and radioactive materials from thorium during the chemical processing after irradiation, as revealed by a literature survey.
Kondo, Ryoichi; Yamamoto, Akio*; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 63(2), p.142 - 153, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The flux distribution tallies using proper orthogonal decomposition, called POD tallies, have been developed to address issues of high-fidelity Monte Carlo simulations. The POD tallies can reduce both dimensionality and statistical error. The present study verifies the applicability of the POD tallies to sub-pin level flux distribution in the two-dimensional C5G7 benchmark. Statistical error estimation is also proposed by applying the circular block bootstrap method to the POD tallies to estimate the statistical error of the flux distribution in a single Monte Carlo calculation. In the verification, the dimensionality of the finely discretized distribution is reduced by more than 90% compared with conventional cell tallies. The statistical error is reduced by more than half as the average value of all tally regions. The proposed approach is confirmed to properly estimate the statistical error of flux distribution considering both the inter-cycle correlation and the correlation between the expansion coefficients of different POD orders.
Gunji, Satoshi; Araki, Shohei; Yoshikawa, Tomoki
Journal of Nuclear Science and Technology, 63(2), p.207 - 215, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Maruyama, Shuhei; Yamamoto, Akio*; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 63(1), p.31 - 44, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 63(1), p.3 - 18, 2026/01
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 HO 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.
Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Katsuyama, Jinya
Journal of Nuclear Science and Technology, 11 Pages, 2026/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Tada, Kenichi; Aizawa, Naoto*; Fujita, Tatsuya*; Fukushima, Masahiro; Pyeon, C. H.*
Journal of Nuclear Science and Technology, 63(1), p.1 - 2, 2026/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This document is the preface to "Special Issue on Progressive Reactor Physics for Current and Future Challenges" published in the Journal of Nuclear Science and Technology.
Mihara, Takeshi; Udagawa, Yutaka
Journal of Nuclear Science and Technology, 17 Pages, 2026/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Fukuda, Kodai; Shiba, Shigeki*; Iwahashi, Daiki*; Gunji, Satoshi
Journal of Nuclear Science and Technology, 14 Pages, 2026/00
Times Cited Count：0Ono, Ayako; Okawa, Tomio*; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 62(12), p.1231 - 1239, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The development of a precise and reliable prediction method for a departure from nucleate boiling (DNB) is urgently needed to design and develop new-generation reactors that enable us to establish a carbon-neutral society. In this study, we consider that the formation of a large vapor mass on the heating surface is the primary trigger of the DNB. Therefore, we embark on the development of a new model for vapor mass formation under forced convective boiling. We assume that the primary bubbles coalesced with each other, which are generated from the nucleation sites, and the coalesced bubble is formed. The nucleation sites are assumed to be distributed on the basis of the Poisson distribution. The large vapor mass is assumed to be formed by the merging of coalesced bubbles when the diameter of the coalesced bubble satisfies the criteria of the slug formation by Mishima. The analysis using the experimental data showed that the proposed model predicted the heat flux to form the large vapor mass well.
Fukuda, Takanari; Yamashita, Susumu; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 62(12), p.1264 - 1278, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study compared three interface capturing schemes (ICSs) for multi-phase flow simulations based on the VOF method, focusing on bubble volume conservation. The THINC/WLIC scheme showed significant VOF diffusion and underestimated total bubble volume, while the original THINC and PLIC conserved bubble volumes. Moreover, an analysis of THINC/WLIC based on a new visualization approach revealed that VOF fragments were ripped off by shear forces around interface, making it unsuitable for accurate void fraction prediction in boiling water reactors. The original THINC may be a viable alternative to PLIC due to its simplicity.
Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya
Journal of Nuclear Science and Technology, 62(12), p.1226 - 1230, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Takei, Hayanori
Journal of Nuclear Science and Technology, 62(11), p.1051 - 1070, 2025/11
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The Japan Atomic Energy Agency is working on the research and development of an accelerator-driven nuclear transmutation system (ADS) for transmuting minor actinides. This system combines a subcritical nuclear reactor with a high-power superconducting proton linear accelerator (JADS-linac). One of the factors limiting the advancement of the JADS-linac is beam trips, which often induce thermal cycle fatigue, thereby damaging the components in the subcritical core. The average beam current of the JADS-linac is 32 times higher than that of the linear accelerator (linac) of the Japan Proton Accelerator Research Complex (J-PARC). Therefore, according to the development stage, comparing the beam trip frequency of the JADS-linac with the allowable beam trip frequency (ABTF) is necessary. Herein the beam trip frequency of the JADS-linac was estimated through a Monte Carlo program using the reliability functions based on the operational data of the J-PARC linac. The Monte Carlo program afforded the distribution of the beam trip duration, which cannot be obtained using traditional analytical methods. Results show that the frequency of the beam trips with a duration exceeding 5 min must be reduced to 27% of the current J-PARC linac level to be below the ABTF.
