Evaluation of Co inventory in the core of the Fukushima Daiichi Nuclear Power Plant; Contribution of fuel deposits to the reactor core inventory

Uchida, Shunsuke*; Kino, Chiaki*; Karasawa, Hidetoshi; Takahatake, Yoko; Koma, Yoshikazu

Journal of Nuclear Science and Technology, 17 Pages, 2025/07

https://doi.org/10.1080/00223131.2025.2499187

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.

Evaluations on the thermal properties of Cs-B-O compounds using density functional theory and phonon vibration calculations

Suzuki, Chikashi

Journal of Nuclear Science and Technology, 62(6), p.542 - 551, 2025/06

https://doi.org/10.1080/00223131.2025.2451824

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 CsO 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 CsBO 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.

Numerical analysis of natural convective heat transfer with porous medium using JUPITER

Uesawa, Shinichiro; Yamashita, Susumu; Sano, Yoshihiko*; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 62(6), p.523 - 541, 2025/06

https://doi.org/10.1080/00223131.2025.2451023

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.

Improvement in automated particle measurement using micromanipulation and large geometry secondary ion mass spectrometry to remove the particle mixing effect of uranium particles

Tomita, Ryohei; Tomita, Jumpei; Suzuki, Daisuke; Miyamoto, Yutaka; Yasuda, Kenichiro

Journal of Nuclear Science and Technology, 10 Pages, 2025/05

https://www.tandfonline.com/doi/full/10.1080/00223131.2025.2512812

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.

Impact of nuclear data updates from JENDL-4.0 to JENDL-5 on burnup calculations of light-water reactor fuels

Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 16 Pages, 2025/04

https://doi.org/10.1080/00223131.2025.2486798

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.

Estimation of influence of implicit effect due to multi-group cross-section perturbations on uncertainty analysis in PWR-UO and -MOX lattice calculations

Fujita, Tatsuya

Journal of Nuclear Science and Technology, 9 Pages, 2025/03

https://doi.org/10.1080/00223131.2025.2470234

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.

Development of neutron self-indication thermometry at J-PARC

Segawa, Mariko; Toh, Yosuke; Maeda, Makoto; Kai, Tetsuya

Journal of Nuclear Science and Technology, 62(3), p.268 - 277, 2025/03

https://doi.org/10.1080/00223131.2024.2424519

Extraction behaviors of minor actinides and rare earth elements with NTA amide extractants

Suzuki, Hideya*; Ban, Yasutoshi

Journal of Nuclear Science and Technology, 62(2), p.157 - 166, 2025/02

https://doi.org/10.1080/00223131.2024.2402366

Chemical interaction of CsOH vapor with UO and Fe-Zr melt

Nakajima, Kunihisa; Takano, Masahide

Journal of Nuclear Science and Technology, 62(1), p.78 - 85, 2025/01

https://doi.org/10.1080/00223131.2024.2389153

At TEPCO's Fukushima Daiichi Nuclear Power Station, it is estimated that considerable amounts of cesium still remain in the reactors from the analysis results using the severe accident analysis codes and the reverse analysis from contaminated water. Since cesium is known to form stable compounds with uranium and zirconium, chemisorption experiments with uranium dioxide pellets and iron-zirconium melts for cesium hydroxide vapor were carried out. As the results, formations of cesium uranate, CsUO, and cesium zirconate, CsZrO, were confirmed, indicating that cesium was chemisorbed on both of the uranium dioxide pellets and the iron-zirconium melts in an Ar-H-HO flow and an Ar-H flow, respectively. Therefore, it was considered that cesium released from fuel might be trapped by chemisorption with fuels and/or iron-zirconium melts during light water reactor severe accidents.

Development of a dissolution method for analyzing the elemental composition of fuel debris using sodium peroxide fusion technique

Nakamura, Satoshi; Ishii, Sho*; Kato, Hitoshi*; Ban, Yasutoshi; Hiruta, Kenta; Yoshida, Takuya; Uehara, Hiroyuki; Obata, Hiroki; Kimura, Yasuhiko; Takano, Masahide

Journal of Nuclear Science and Technology, 62(1), p.56 - 64, 2025/01

https://doi.org/10.1080/00223131.2024.2381553

A dissolution method for analyzing the elemental composition of fuel debris using the sodium peroxide (NaO) fusion technique has been developed. Herein, two different types of simulated debris materials (such as solid solution of (Zr,RE)O and molten core-concrete interaction products (MCCI)) were taken. At various temperatures, these debris materials were subsequently fused with NaO in crucibles, which are made of different materials, such as Ni, AlO, Fe, and Zr. Then, the fused samples are dissolved in nitric acid. Furthermore, the effects of the experimental conditions on the elemental composition analysis were evaluated using inductively coupled plasma-atomic emission spectroscopy (ICP-AES), which suggested the use of a Ni crucible at 923 K as an optimum testing condition. The optimum testing condition was then applied to the demonstration tests with Three Mile Island unit-2 (TMI-2) debris in a shielded concrete cell, thereby achieving complete dissolution of the debris. The elemental composition of TMI-2 debris revealed by the proposed dissolution method has good reproducibility and has an insignificant contradiction in the mass balance of the sample. Therefore, this newly developed reproducible dissolution method can be effectively utilized in practical applications by dissolving fuel debris and estimating its elemental composition.