Temperature-dependent deformation behavior of dual-phase medium-entropy alloy; In-situ neutron diffraction study

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

https://doi.org/10.1016/j.jmst.2024.11.057

Numerical investigation of the accuracy of a conductance-type wire-mesh sensor for a single spherical bubble and bubbly flow

Uesawa, Shinichiro; Ono, Ayako; Nagatake, Taku; Yamashita, Susumu; Yoshida, Hiroyuki

Journal of Nuclear Science and Technology, 62(5), p.432 - 456, 2025/05

https://doi.org/10.1080/00223131.2024.2447366

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.

Development of scaling parameter S 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

https://doi.org/10.1115/1.4067484

Measurements of neutron capture cross-section for nuclides of interest in decommissioning (II); Fe(n,)Fe

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

Journal of Nuclear Science and Technology, 62(3), p.300 - 307, 2025/03

https://doi.org/10.1080/00223131.2024.2427244

Verification of direct coupling code system using FRENDY version 2 and GENESIS for light water reactor lattices

Fujita, Tatsuya; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 62(2), p.179 - 196, 2025/02

https://doi.org/10.1080/00223131.2024.2410918

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.

Synchrotron radiation photoemission electron microscopy study on radioactive cesium-bearing microparticle collected in Fukushima

Yoshigoe, Akitaka; Tsuda, Yasutaka; Kobata, Masaaki; Okane, Tetsuo; Satou, Yukihiko; Okochi, Takuo*

e-Journal of Surface Science and Nanotechnology (Internet), 23(1), p.16 - 21, 2025/02

https://doi.org/10.1380/ejssnt.2025-004

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.

Measurements of neutron capture cross-section for nuclides of interest in decommissioning (III); Er(n,)Er and Hf(n,)Hf reactions

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

Journal of Nuclear Science and Technology, 14 Pages, 2025/00

https://doi.org/10.1080/00223131.2025.2457589

Effect of sea salt on the interaction between simulated corium and concrete

Sudo, Ayako; Sato, Takumi; Takano, Masahide

Journal of Nuclear Science and Technology, 9 Pages, 2025/00

https://doi.org/10.1080/00223131.2025.2456565

During the progression of the severe accident at the Fukushima Daiichi Nuclear Power Station, seawater flowed down and was predicted to react with molten corium and concrete. For the removal and storage of fuel debris, knowing the effects of seawater components on the characteristics of reaction products in the fuel debris is crucial. To understand changes in the microstructure of fuel debris, a reaction test was conducted by introducing sea salt to simulated corium and concrete under a temperature gradient. Among the components of sea salt, sulfur formed iron sulfide during metallic precipitation. Analysis of vaporized species indicated that most of Cl, some Na and K in the sea salt might volatilize during heating rather than react with simulated corium and concrete. Calcium and a small amount of Mg, Na, and K in the sea salt might be trapped in the silicate glass.

Integrity evaluation of boundary function of main components in nuclear plants during severe accidents

Tsukimori, Kazuyuki; Yada, Hiroki

Journal of Pressure Vessel Technology, 147, p.031901_1 - 031901_9, 2025/00

https://doi.org/10.1115/1.4067760

After the accident at the Fukushima Daiichi Nuclear Power Plant, very strict safety measures were implemented for nuclear power plants in Japan. It thus becomes a crucial issue if the safety of a plant is maintained or not at beyond design basis events. In this study, head plates and bellows were examined as components that compose the parts of the boundary of vessels that contain the primary coolant of a prototype fast breeder reactor. The behaviors of buckling, post-buckling deformation, and penetration failure, that is, loss of boundary function of these components with increasing pressure were investigated. The series of this research program started in FY2013 and the research proceeded step by step. The new result in this paper is the application of the proposed criteria to head plates and bellows, and a conservative estimation of penetration failure of these components is obtained.