Development of 1D-CFD coupling method for natural circulation analyses through benchmark analyses of shutdown heat removal tests in EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Fujisaki, Tatsuya*; Murakami, Satoshi*

Annals of Nuclear Energy, 226, p.111896_1 - 111896_11, 2026/02

https://doi.org/10.1016/j.anucene.2025.111896

At the Japan Atomic Energy Agency, a multilevel simulation (MLS) methodology which enables consistent evaluation from whole plant behavior to local phenomena in the plant components is being developed to attempt plant design and enhance the safety of sodium-cooled fast reactors. To validate the coupling method in the MLS system, the 1D-CFD coupling method using Super-COPD for 1D plant dynamics analysis and Fluent for multi-dimensional CFD analysis was applied to the analyses of loss of flow tests in EBR-II. It was confirmed that it could predict multi-dimensional thermal-hydraulic phenomena such as thermal stratification in the upper plenum, Z-shaped pipe, and cold pool, holding the whole plant behavior simultaneously. Moreover, the applicability of the 1D-CFD coupling method to the evaluation of the phenomena in natural circulation conditions was confirmed by comparing the results of the 1D-CFD couple analyses and the measured data.

Validation of underwater radiation monitoring detector

Ji, W.*; Lee, E.*; Ji, Y.-Y.*; Ochi, Kotaro; Yoshimura, Kazuya; Funaki, Hironori; Sanada, Yukihisa

Nuclear Engineering and Technology, 58(2), p.103933_1 - 103933_6, 2026/02

https://doi.org/10.1016/j.net.2025.103933

We aimed to validate the performance of an in situ underwater radiation detector, MARK-U1 (Monitoring of Ambient Radiation of KAERI - Underwater), was used to estimate Cs activity concentration in river and reservoir sediment at predicted sites of contamination. Additionally, underwater core samples were collected to measure the radioactivity using a high-purity germanium (HPGe) detector. To estimate radioactivity, a conversion factor was derived by comparing the measured spectrum and Cs activity in the sample. A Monte Carlo N-Particle (MCNP) simulation was conducted to determine the effective source geometry for in situ measurement. The simulation results correlated well with the on-site MARK-U1 monitoring results, with a deviation of 31.62%. These findings validate the performance of the in situ detector. This device can therefore be used to estimate Cs activity concentration in the underwater sediment via on-site monitoring, without requiring sample collection.

Random media criticality analysis using randomized Fourier series and incomplete randomized Weierstrass function

Ueki, Taro

Progress in Nuclear Energy, 191, p.106007_1 - 106007_11, 2026/01

https://doi.org/10.1016/j.pnucene.2025.106007

Criticality analysis of continuously mixed random media is crucial for safely retrieving fuel debris. Initially, a Monte Carlo method was established using the Incomplete Randomized Weierstrass Function (IRWF) to model a single-mode inverse power law power spectrum. However, image analysis showed that oxide debris mock-ups require a more complex model. To address this, a new function called the Randomized Fourier Series (RFS) was developed to represent arbitrary power spectra. RFS is versatile, incorporating Brownian motion models and aiding reactor physicists in analyzing various scenarios. Numerical results compare the fluctuation of neutron multiplication factor in various media generated by RFS and IRWF, identifying the spectral range most affecting k.

Rational physical protection design of transuranium fuel cycle site with accelerator-driven system by using material attractiveness

Oizumi, Akito; Sagara, Hiroshi*

Annals of Nuclear Energy, 223, p.111677_1 - 111677_12, 2025/12

https://doi.org/10.1016/j.anucene.2025.111677

This study aims to provide a new rational physical protection (PP) design method by using () and to design a rational PP system for a site of the transuranium fuel cycle with accelerator-drive systems (ADSs cycle) using the new method. First, the new rational PP design method with different PP design requirements for each was generalized based on the definitions of a national standard method defined by the US Department of Energy, the joint US-Japan study, and the International Atomic Energy Agency. A new PP categorization of Uranium (U), including U-234, which is abundant in the ADS cycle, was also developed based on . Second, a PP design was conducted for a general BWR site with MOX fuel and the ADS cycle site by using the new rational method. It was clarified that the highest overall of the items within the ADS cycle site was lower than that of the MOX fuel assembly within the BWR site. The BWR site was determined to be Category I requiring the inner area. The PP design requirement level of the ADS cycle site was determined to be Category II, which does not require an inner area, while the ADS cycle site would have been classified as Category I if the PP design had been conducted using the conventional method.

A Methodology for the design of non-uniform core configurations in the modified STACY facility

Dechenaux, B.*; Brovchenko, M.*; Araki, Shohei; Gunji, Satoshi; Suyama, Kenya

Annals of Nuclear Energy, 223, p.111555_1 - 111555_11, 2025/12

https://doi.org/10.1016/j.anucene.2025.111555

Flow regimes and heat transfer for opposing flow mixed convection in the thermal entry region of a vertical tube

Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*

International Journal of Heat and Mass Transfer, 252, p.127451_1 - 127451_16, 2025/12

https://doi.org/10.1016/j.ijheatmasstransfer.2025.127451

Improvement of a reinforcement learning model to assist operators in controlling a nuclear power plant under abnormal operating conditions

Yoshikawa, Masanori; Seki, Akiyuki*; Okita, Shoichiro; Takaya, Shigeru; Yan, X.

Nuclear Engineering and Design, 444, p.114350_1 - 114350_9, 2025/12

https://doi.org/10.1016/j.nucengdes.2025.114350

Effect of grain refinement on cracks occurring in SUS304L stainless steel under nuclear reactor operating conditions

Hirota, Noriaki; Takeda, Ryoma; Ide, Hiroshi; Tsuchiya, Kunihiko; Kobayashi, Yoshinao*

Nuclear Materials and Energy (Internet), 45, p.102009_1 - 402009_10, 2025/12

https://doi.org/10.1016/j.nme.2025.102009

Using SUS304L stainless steel, which is employed in reactor structural components, the effects of grain refinement on stress corrosion cracking occurring under nuclear reactor operating conditions were investigated. As a result, after conducting slow strain rate testing (SSRT) in air and nuclear reactor operating environments, a comparison of the tensile properties of SUS304L with the same grain size revealed that elongation significantly decreased with increasing grain size under nuclear reactor operating conditions. In SSRT conducted in air, the -value obtained from the Hall-Petch relationship was lower than the conventional values. Observations showed the absence of cracks on SUS304L with 0.59 and 1.52 m grains; however, SUS304L with larger grains exhibited rougher fracture surfaces and side cracks. Thin oxide films were formed on SUS304L with 0.59 m and 1.52 m grains, while SUS304L with coarse grains of 28.4 m or larger enabled the formation of oxide films with over 2 m thickness. CrO films were formed on SUS304L with 0.59 m, 1.52 m, and 28.4 m, while CrO and Fe based oxides were formed on SUS304L with 39.5 m and 68.6 m. Crystal orientation analysis revealed linear surface layers without cracks in the -phase for SUS304L with 0.59 m and 1.52 m. In materials with Larger grain sizes, surface irregularities and cracks were observed in the -phase.In fine-grained SUS304L, lattice diffusion caused uniform O diffusion in the -phase, resulting in the formation of a thin CrO layer that suppressed cracks. In coarse-grained SUS304L, grain boundary diffusion caused Fe oxide formation at the grain boundaries, weakening them, and supersaturated O led to the formation of thick films comprising CrO and Fe-based oxides, resulting in peeling and cracking.

In situ reaction and alteration of mudstone-cement interfaces at the Horonobe URL; Comparison between ordinary Portland cement and low alkaline cement

Dei, Shuntaro; Shibata, Masahito*; Negishi, Kumi*; Sugiura, Yuki; Amano, Yuki; Bateman, K.*; Wilson, J.*; Yokoyama, Tatsunori; Kagami, Saya; Takeda, Masaki; et al.

Results in Earth Sciences (Internet), 3, p.100097_1 - 100097_16, 2025/12

https://doi.org/10.1016/j.rines.2025.100097

Interactions between cement and host rock in geological repositories for radioactive waste will result in a chemically disturbed zone, which may potentially affect the long-term safety. This paper investigates the chemical evolution at the interface between cement (Ordinary Portland Cement: OPC and Low Alkaline Cement: LAC) and mudstone after 11 years of in situ reactions at the Horonobe Underground Research Laboratory. The study combines various analytical techniques to identify the key reactions at the cement-rock interface, including cement dissolution, precipitation of secondary minerals such as calcite and C-(A-)S-H phases, cation exchange in montmorillonite and reduced porosity in mudstone. The study also highlights the effects of cement-mudstone interactions on radionuclide migration, such as reduction of diffusivity due to reduced porosity and enhancement of sorption due to incorporation into secondary minerals in the altered mudstone.

Investigation of carbonation depth in concrete of Waste Treatment Facility No. 1 and Waste Treatment Facility No. 2

Asahi, Miyako; Hozumi, Shinsei; Suzuki, Ichiro; Seya, Manato; Mori, Masakazu; Sakamoto, Yu; Kinoshita, Junichi; Sudo, Tomoyuki

JAEA-Testing 2025-005, 15 Pages, 2025/11

JAEA-Testing-2025-005.pdf:1.37MB

Radioactive wastes generated from the nuclear facilities of the Nuclear Science Research Institute is collected at Radioactive Waste Treatment Facilities, where it is treated or stored according to its radioactivity level and properties. Radioactive Waste Treatment Facilities is composed of several buildings, of which Waste Treatment Facility No.1 has the incinerator, and Waste Treatment Facility No.2 has the equipment treating solid waste in concrete cells. Since Waste Treatment Facility No.1 and No.2 were built more than 40 years ago, the carbonation test was conducted in FY2024 to consider future maintenance management. In addition, the progress of carbonation was predicted based on the test results. This report describes an overview of the carbonation test and the prediction of the progress of carbonation.