Cold sintering of whitlockite-based phosphate ceramics for ALPS carbonate slurry waste immobilization

Gubarevich, A.*; Kubo, Ryotaro*; Arisaka, Makoto; Osugi, Takeshi; Yoshida, Katsumi*; Takeshita, Kenji*

Journal of Nuclear Science and Technology, 63(6), p.667 - 676, 2026/06

https://doi.org/10.1080/00223131.2025.2592963

To immobilize and solidify carbonate-based Advanced Liquid Processing System (ALPS) sediment wastes, this study introduces a chemical transformation process that converts these wastes into calcium and magnesium phosphate phases, followed by densification using a novel cold sintering press (CSP) technique. Simulated calcium-magnesium carbonate slurries were treated with phosphoric acid to synthesize calcium and magnesium phosphates, which were then sintered at 300-500C using CSP. The effects of the calcium-to-magnesium ratio, strontium incorporation, and sodium chloride addition on phase composition and CSP densification were investigated. Dense bulk samples were successfully fabricated and characterized using X-ray diffraction, scanning electron microscopy, energy-dispersive X-ray spectroscopy, and the Archimedes method. The results showed that the chemical transformation process led to the formation of whitlockite and newberyite, with the calcium-to-magnesium ratio determining the relative proportions of these phases. Strontium was effectively incorporated into the whitlockite crystal structure, while newberyite enhanced densification through a dehydration-driven process. Sodium chloride had no effect on chemical transformation and was not found in the final solid product. These results show that direct conversion of calcium-magnesium carbonate slurries to whitlockite-based phosphate ceramics, followed by CSP, enables stable solidification, making this method promising for ALPS sediment waste management.

Measurements of neutron capture cross-sections for nuclides of interest in decommissioning (V); Zr(n, )Zr and Zr(n, )Zr reactions at JRR-3

Nakamura, Shoji; Kimura, Atsushi; Endo, Shunsuke; Rovira Leveroni, G.; Shibahara, Yuji*

Journal of Nuclear Science and Technology, 63(6), p.653 - 666, 2026/06

https://doi.org/10.1080/00223131.2025.2586632

Performance evaluation for rapid-dose estimation of radioactive plume dispersion based on pre-simulation database of wind conditions by large-eddy simulation

Sato, Takuto; Nakayama, Hiromasa; Satoh, Daiki

Journal of Nuclear Science and Technology, 63(4), p.426 - 442, 2026/04

https://doi.org/10.1080/00223131.2025.2555784

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.

Simple technique for the preparation of uranium-impregnated porous silica particles and their application as working standard particles for analysis of the safeguards environmental samples

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

Journal of Nuclear Science and Technology, 63(4), p.443 - 454, 2026/04

https://doi.org/10.1080/00223131.2025.2557376

CFD simulation of the XR2-1 experiment with the JUPITER code

Yamashita, Susumu; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 19 Pages, 2026/04

https://doi.org/10.1080/00223131.2026.2664459

This study investigates the applicability of the mechanistic CFD code JUPITER to three-dimensional melt relocation phenomena in nuclear reactor cores during severe accidents. The XR2-1 BWR metallic melt relocation experiment was analyzed as an integral-effect test case focusing exclusively on melt relocation behavior under an inert atmosphere. A detailed three-dimensional model of the XR2-1 test section, including fuel assemblies, control blades, and lower core support structures, was constructed, and time-dependent injections of SS/B4C and Zircaloy melts were simulated under experimentally based thermal conditions. The simulation results were evaluated through qualitative comparisons of melt relocation paths and quantitative comparisons of relocated material volumes in key regions of the test section. The analysis successfully reproduced the three experimentally observed melt relocation paths - through the control blade guide tube, nosepiece and inlet nozzle, and along the channel box region - without the formation of internal blockages. Quantitative comparisons showed reasonable agreement with post-test X-ray tomographic measurements for most evaluation regions. These results demonstrate that JUPITER can realistically capture three-dimensional melt relocation behavior in complex core geometries and indicate its potential usefulness as a complementary tool to large-scale experiments for evaluating severe accident melt relocation phenomena. This study identifies major problems in analyzing fuel assembly melt relocation behavior using the current JUPITER code, including the excessive computational cost of radiative heat - transfer calculation, insufficient computational grid resolution, and the absence of fluid-structure interaction modeling.

Experimental measurements for the first series of the modified STACY critical assembly with simple core configurations and experimental analysis using simplified computational models

Gunji, Satoshi; Araki, Shohei; Yoshikawa, Tomoki

Journal of Nuclear Science and Technology, 63(2), p.207 - 215, 2026/02

https://doi.org/10.1080/00223131.2025.2592962

Feasibility study on the production of Th as a long-life Ac generator using the experimental fast reactor Joyo

Sasaki, Yuto; Maeda, Shigetaka; Fukasawa, Tetsuo*; Takaki, Naoyuki*

Journal of Nuclear Science and Technology, 63(2), p.154 - 165, 2026/02

https://doi.org/10.1080/00223131.2025.2558244

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.

Burnup calculation using POD-based neutron spectrum reconstruction

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

https://doi.org/10.1080/00223131.2025.2560559

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.

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, 63(1), p.3 - 18, 2026/01

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.

A Finite element analysis study on the fracture pattern change of fuel cladding under PCMI loading conditions

Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Katsuyama, Jinya

Journal of Nuclear Science and Technology, 11 Pages, 2026/00

https://doi.org/10.1080/00223131.2025.2611411