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Gu, G. H.*; Jeong, S. G.*; Heo, Y.-U.*; Harjo, S.; Gong, W.; Cho, J.*; Kim, H. S.*; 4 of others*
Journal of Materials Science & Technology, 223, p.308 - 324, 2025/07
Times Cited Count：0 Percentile：75.81(Materials Science, Multidisciplinary)
Co inventory in the core of the Fukushima Daiichi Nuclear Power Plant; Contribution of fuel deposits to the reactor core inventoryUchida, Shunsuke*; Kino, Chiaki*; Karasawa, Hidetoshi; Takahatake, Yoko; Koma, Yoshikazu
Journal of Nuclear Science and Technology, 17 Pages, 2025/07
Evaluation of radioactive nuclide behavior on and after the accident is important for the estimating radioactive nuclide composition in the wastes. The reactor core inventories have been obtained from the ORIGEN2 analysis, but the inventory of activation products is determined by the amount of their parent nuclides which are impurities contained in the structural materials. The ORIGEN2 does not treat fuel deposits including the impurities. Estimation of the initial Co-60 inventory in accurate is needed on the evaluation of some kinds of radioactive nuclide amount, since it is possible Co-60 is standard in the scaling factor. In this study, contribution of fuel deposits to the reactor core inventory was estimated by comparing the amount of Co-60 and Ni-63 calculated by the amounts of deposition by the microlayer-evaporation and drying-out model (MEDO model) and the result of the ORIGEN2 analysis, and then the method of estimating the reactor core inventory was proposed.
Ho(n,
)
Ho reactionsNakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Rovira Leveroni, G.; Kimura, Atsushi
Journal of Nuclear Science and Technology, 14 Pages, 2025/07
Murakami, Masashi; Yoshida, Yukihiko; Nango, Nobuhito*; Kubota, Shogo*; Kurosawa, Takuya*; Sasaki, Toshiki
Journal of Nuclear Science and Technology, 62(7), p.650 - 661, 2025/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Hayakawa, Kosetsu*; Muraguchi, Masakazu*; Masebo, Yumeno*; Kojima, Yoichiro*; Oda, Masato*; Yabuta, Rina*; Ishii, Hiroyuki*; Honda, Mitsunori
e-Journal of Surface Science and Nanotechnology (Internet), p.2025-041_1 - 2025-041_6, 2025/07
We are developing materials using clay minerals for thermoelectric materials that can be used at high temperatures in environmental friendly applications. We have succeeded in obtaining thermoelectric properties at high temperatures, but we do not know why these properties are exhibited. In order to elucidate the structural changes of weathered biotite under high temperature conditions, we will develop an XAFS analysis method and clarify the local structure of Fe at high temperatures. The local structure analysis of Fe in samples sintered in atmospheric and vacuum environments reveals changes in the chemical bonding state of Fe.
Takei, Hayanori
Journal of Nuclear Science and Technology, 45 Pages, 2025/06
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.
Suzuki, Chikashi
Journal of Nuclear Science and Technology, 62(6), p.542 - 551, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Boron (B) can chemically react with cesium (Cs) to form Cs-B-O compounds and affect the chemical behavior of Cs during severe accidents. The author evaluated the thermal properties of solid Cs-B-O compounds using density functional theory and phonon vibration calculations with Cs
O and BO
as solid reference materials. These evaluations indicate that the calculated thermal properties are in good agreement with the reported ones. The author calculated the reaction enthalpy and Gibbs free energy values of solid CsBO from CsO and BO to obtain fundamental data in solid systems. The deviations between the calculation and the reported data in this study are comparable with those in the previous study. The author estimated the Gibbs free energy values of the CsB
O
reaction from CsBO and boric acid. The differences between these estimations and those in the recent investigation are within the error of the reaction energies in the experiments.
Uesawa, Shinichiro; Yamashita, Susumu; Sano, Yoshihiko*; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 62(6), p.523 - 541, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) has developed a numerical method with the JUPITER code with a porous medium model to calculate the thermal behavior in PCVs of 1F. In this study, we performed an experiment and numerical simulation of the natural convective heat transfer with the porous medium to validate JUPITER with the porous medium model. In comparison of the temperature and velocity distributions between the experiment and simulation, the temperature distribution in the simulation was in good agreement with the distribution in the experiment except the temperature near the top surface of the porous medium. The velocity distribution also agreed qualitatively with the experimental result. In addition, we also performed the numerical simulations with various effective thermal conductivity models to discuss the effect of the conductivity based on the internal structure of porous media on the natural convective heat transfer. The result indicated that the temperature distribution in the porous medium and the velocity distribution of the natural convection were significantly different for each model, and thus the conductivity of the fuel debris was one of the key parameters of in the thermal behavior analysis in 1F.
Hasegawa, Kunio; Yamaguchi, Yoshihito; Udyawar, A.*
Journal of Pressure Vessel Technology, 147(3), p.034501_1 - 034501_7, 2025/06
Times Cited Count：0 Percentile：0.00(Engineering, Mechanical)Katsumi, Toshiyuki*; Zemba, Atsushi*; Kogishi, Makoto*; Iwanaga, Kohei*; Abe, Satoshi; Di
vart, P.*; Chaumeix, N.*; Kadowaki, Satoshi*
Journal of Thermal Science and Technology (Internet), 20(1), p.25-00103_1 - 25-00103_13, 2025/06
Times Cited Count：0 Percentile：0.00In the serious accident at Fukushima Daiichi Nuclear Power Station, the presence of steam together with flammable organic compounds affected the hydrogen explosion. To investigate the effects of addition of inert gas, i.e. steam or nitrogen, on the explosion characteristics, we conducted the experiments of spherically expanding hydrogen-methane-air premixed flames in closed combustion vessels. Two types of vessels were used, and expanding flames were caught by Schlieren method. The flame propagation velocity depending on the flame radius was obtained by analyzing Schlieren images. When the flame radius was sufficiently small, smooth surface was found. The addition of inert gas to hydrogen-methane-air premixtures caused the decrease of propagation velocity of unstretched flame. When the flame radius was large, on the other hand, cellular surface generated by intrinsic instability was found. In this range, the flame acceleration was confirmed, which was induced by the evolution of cellular surface. We obtained the parameters of flame acceleration model and predicted the flame propagation velocity depending on the flame radius. The increment coefficient normalized by the propagation velocity of unstretched flame became larger at low equivalence ratios, which was due to stronger diffusive-thermal instability. Under the same equivalence ratio, the inert gas addition caused the increase of normalized increment coefficient. This denoted that the inert gas addition promoted the instability of premixed flames, which was due to the reduction of the effective Lewis number. The maximum pressure in a combustion vessel became lower in the case of inert gas addition. Moreover, the maximum pressure of experiments was lower than that of calculations under the adiabatic conditions, because of heat loss during premixed combustion. The obtained results were valuable knowledge to elucidate the hydrogen explosion at Fukushima Daiichi Nuclear Power Station.
Tomita, Ryohei; Tomita, Jumpei; Suzuki, Daisuke; Miyamoto, Yutaka; Yasuda, Kenichiro
Journal of Nuclear Science and Technology, 10 Pages, 2025/05
A new automated particle measurement (APM) combined with micromanipulation using large geometry secondary ion mass spectrometry instrument was proposed and demonstrated to remove the particle mixing effect, which indicated that the aggregation of uranium particles was detected as a single uranium particle, from APM results. The results showed that the new APM method was more effective than the traditional APM method in removing the particle mixing effect from the APM results and determining the existence of minor uranium isotopes in the samples.
Sonehara, Masateru; Okano, Yasushi; Uchibori, Akihiro; Oki, Hiroshi*
Journal of Nuclear Science and Technology, 62(5), p.403 - 414, 2025/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)For sodium-cooled fast reactors, understanding sodium combustion behaviour is crucial for managing sodium leakage accidents. In this study, we perform benchmark analyses of the Sandia National Laboratories (SNL) T3 experiment using the multi-dimensional thermal hydraulic code AQUA-SF. Conducted in an enclosed space with a large vessel volume of 100 m
and a sodium mass flow rate of 1 kg/s, the experiment highlighted the multi-dimensional effects of local temperature increase shortly after sodium injection. This study aims to extend the capabilities of AQUA-SF by focusing on the simulation of these multi-dimensional temperature variations, in particular the formation of high temperature regions at the bottom of the vessel. The proposed models include the temporary stopping of sodium droplet ignition and spray combustion of sodium splash on the floor. Furthermore, it has been shown that additional heat source near the floor is essential to enhance the reproduction of the high temperature region at the bottom. Therefore, case studies including sensitivity analyses of spray cone angle and prolonged combustion of droplets on the floor are conducted. This comprehensive approach provides valuable insights into the dynamics of sodium combustion and safety measures in sodium-cooled fast reactors.
Uesawa, Shinichiro; Ono, Ayako; Nagatake, Taku; Yamashita, Susumu; Yoshida, Hiroyuki
Journal of Nuclear Science and Technology, 62(5), p.432 - 456, 2025/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)We performed electrostatic simulations of a wire-mesh sensor (WMS) for a single spherical bubble and bubbly flow to clarify the accuracy of the WMS. The electrostatic simulation for the single bubble showed the electric current density distribution and the electric current path from the excited transmitter to receivers for various bubble locations. It indicated systematic errors based on the nonuniform current density distribution around the WMS. The electrostatic simulation for the bubbly flow calculated by the computational fluid dynamics code, JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER), indicated that the WMS had difficulty in quantitatively measuring the intermediate values of the instantaneous void fraction between 0 and 1 because they cannot be estimated by previous transformation methods from the WMS signal to the void fraction, such as linear approximation or Maxwell's equation, and have a significant deviation of the void fraction of 
0.2 for the WMS signal. However, the electrostatic simulation indicated that the time-averaged void fractions around the center of the flow channel can be estimated using linear approximation, and the time-averaged void fraction near the wall of the flow channel can be estimated using Maxwell's equation.
Sato, Yuki; Terasaka, Yuta; Ichiba, Yuta*
Journal of Nuclear Science and Technology, 62(4), p.389 - 400, 2025/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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 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.
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) and -MOX lattice calculationsFujita, Tatsuya
Journal of Nuclear Science and Technology, 9 Pages, 2025/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study estimated the influence of implicit effect on the k-infinity uncertainty in the PWR-UO and -MOX fuel lattice geometries. Firstly, the preliminary investigation was performed, where the influence of implicit effect was roughly estimated based on the sandwich formula using the cross-section (XS) covariance matrix and the sensitivity coefficient. It was confirmed that the influence of implicit effect became large in the fission and (n,) reactions of heavy nuclides and the change of this dependence was small for the burnup of UO and MOX fuel assemblies. Then, focussing on the heavy nuclides, the influence of implicit effect was compared under several energy group conditions of the XS covariance matrix and neutron transport calculation. For Pu and 
Pu, the noticeable influence of implicit effect was observed in MOX fuel pin-cell geometry. However, increasing the number of energy groups for neutron transport calculations and that of the XS covariance matrix can reduce the influence of implicit effect. Consequently, by appropriately setting the number of energy groups for neutron transport calculations and that of the XS covariance matrix, it became practically possible not to explicitly consider the implicit effect during the random sampling.
Segawa, Mariko; Toh, Yosuke; Maeda, Makoto; Kai, Tetsuya
Journal of Nuclear Science and Technology, 62(3), p.268 - 277, 2025/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Kaburagi, Masaaki; Miyamoto, Yuta; Mori, Norimasa; Iwai, Hiroki; Tezuka, Masashi; Kurosawa, Shunsuke*; Tagawa, Akihiro; Takasaki, Koji
Journal of Nuclear Science and Technology, 62(3), p.308 - 316, 2025/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)
Fe(n,)
FeNakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Rovira Leveroni, G.; Kimura, Atsushi
Journal of Nuclear Science and Technology, 62(3), p.300 - 307, 2025/03
Times Cited Count：0 Percentile：43.12(Nuclear Science & Technology)