Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Luu, V. N.; Taniguchi, Yoshinori; Udagawa, Yutaka; Tasaki, Yudai; Katsuyama, Jinya
Annals of Nuclear Energy, 230, p.112114_1 - 112114_14, 2026/06
Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Fujisaki, Tatsuya*; Murakami, Satoshi*
Annals of Nuclear Energy, 226, p.111896_1 - 111896_11, 2026/02
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.
Kawaguchi, Munemichi*; Ikeda, Asuka; Saito, Junichi
Annals of Nuclear Energy, 226, p.111880_1 - 111880_9, 2026/02
Times Cited Count：0Rovira Leveroni, G.; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Iwamoto, Osamu; Iwamoto, Nobuyuki; Katabuchi, Tatsuya*
Annals of Nuclear Energy, 225, p.111688_1 - 111688_18, 2026/01
Ueki, Taro
Progress in Nuclear Energy, 191, p.106007_1 - 106007_11, 2026/01
Times Cited Count：0Criticality 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
.
Dechenaux, B.*; Brovchenko, M.*; Araki, Shohei; Gunji, Satoshi; Suyama, Kenya
Annals of Nuclear Energy, 223, p.111555_1 - 111555_11, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Oizumi, Akito; Sagara, Hiroshi*
Annals of Nuclear Energy, 223, p.111677_1 - 111677_12, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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.
Nakase, Masahiko*; Mishima, Ria; Abe, Takumi; Okamura, Tomohiro*; Asano, Hidekazu*
Annals of Nuclear Energy, 224, p.111569_1 - 111569_14, 2025/12
Odaira, Naoya*; Kodama, Katsuaki; Ito, Daisuke*; Saito, Yasushi*; Parker, J. D.*; Shinohara, Takenao
Nuclear Materials and Energy (Internet), 45, p.102005_1 - 102005_7, 2025/12
Hirota, Noriaki; Takeda, Ryoma; Ide, Hiroshi; Tsuchiya, Kunihiko; Kobayashi, Yoshinao*
Nuclear Materials and Energy (Internet), 45, p.102009_1 - 402009_10, 2025/12
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. Cr
O
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.
Zhang, H.*; Umehara, Yutaro*; Horiguchi, Naoki; Yoshida, Hiroyuki; Eto, Atsuro*; Mori, Shoji*
Energy, 335, p.138090_1 - 138090_18, 2025/10
Nuclear power is a key low-carbon energy source for a carbon-neutral future. In boiling water reactors (BWRs), steam-water annular flow near fuel rods is crucial for reactor safety, but its high-temperature, high-pressure conditions (285
C, 7 MPa) make direct measurement challenges. To address this, we used an HFC134a-ethanol system at lower conditions (40C, 0.7 MPa) to simulate BWR annular flow. Using a high-speed camera and the constant electric current method, we analyzed liquid-film characteristics, wave velocity and frequency. We also examined surface tension and interfacial shear stress effects. Furthermore, we proposed a new correlation for base film thickness.
Aoki, Takeshi; Shimizu, Atsushi; Ishii, Katsunori; Morita, Keisuke; Mizuta, Naoki; Kurahayashi, Kaoru; Yasuda, Takanori; Noguchi, Hiroki; Nomoto, Yasunobu; Iigaki, Kazuhiko; et al.
Annals of Nuclear Energy, 220, p.111503_1 - 111503_7, 2025/09
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Aiming to establish coupling technologies between a high temperature gas cooled reactor and a hydrogen production plant, JAEA has initiated the HTTR Heat Application Test Project and is conducting the safety design and the safety analysis for the licensing of the HTTR Heat Application Test Facility. The present study proposed a relative evaluation methodology for the demarcation of applicable laws and design standards for the nuclear hydrogen production system and applied it to the HTTR Heat Application Test Facility. The evaluation results showed that a candidate applying the High Pressure Gas Safety Act to the Heat Application Test Facility (hydrogen production plant) and design standards established under the High Pressure Gas Safety Act to the steam reformer did not show the lowest category in any of the metrics, and was proposed as the most superior demarcation option for the HTTR Heat Application Test Facility.
Soma, Shu; Ishigaki, Masahiro*; Shibamoto, Yasuteru
Annals of Nuclear Energy, 219, p.111455_1 - 111455_12, 2025/09
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Nguyen, H. H.
Annals of Nuclear Energy, 218, p.111361_1 - 111361_9, 2025/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study examined the criticality characteristics of a partially damaged reactor model, in which fuels located at the core center melt into fuel debris of varying shapes, while fuels situated at the core edges remain intact. The investigation was conducted using the Serpent code with the JENDL-5 nuclear data library. The results of the calculations indicate that when the volume of fuel debris is small and maintained at a constant level, the shape of the fuel debris does not result in significant alterations in the variation law of k of the system. In contrast, for the scenario in which the volume of the fuel debris is variable, the k variation law can be divided into two groups for the reference case with a system temperature of 300 K and no boron in the water. The first group comprises fuel debris with shapes that are cuboid and cylindrical, while the second group comprises fuel debris with shapes that are spherical, cone-shaped, and truncated cone-shaped.
Rizaal, M.; Nakajima, Kunihisa; Suzuki, Eriko; Miwa, Shuhei
Annals of Nuclear Energy, 218, p.111433_1 - 111433_10, 2025/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Ito, Ayumi*; Kanno, Tatsuya*; Iwama, Takayuki*; Ueda, Shigeru*; Sato, Takumi; Nagae, Yuji
Annals of Nuclear Energy, 217, p.111333_1 - 111333_14, 2025/07
In the Fukushima Daiichi Nuclear Power Station Unit 2, the formation of a metallic pool, mainly comprising Fe and Zr, has been proposed as a mechanism contributing to the failure of the reactor pressure vessel. This study focuses on material interactions during the early core degradation that led to metallic pool formation in the late phase of the in-vessel degradation process. Initially, two compositions, Fe-87Zr and Fe-15Zr (at%), were heated to the liquidus temperature of 1723 K, dropped onto SS at lower temperatures, and the metallographic structure of the reaction products was examined. Subsequently, the Fe-87Zr melt at temperatures ranging from 1723 to 1873 K was dropped onto oxidized SS to evaluate the influence of the oxide layer on degradation. This study confirmed that the liquidus temperatures of all intermetallic compounds were below 2000 K, and the metallic debris could be a source of the "metallic pool formation" predicted by recent severe accident analysis.
Araki, Shohei; Aizawa, Eiju; Murakami, Takahiko; Arakaki, Yu; Tada, Yuta; Kamikawa, Yutaka; Hasegawa, Kenta; Yoshikawa, Tomoki; Sumiya, Masato; Seki, Masakazu; et al.
Annals of Nuclear Energy, 217, p.111323_1 - 111323_8, 2025/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)JAEA has modified the STACY from a homogeneous system using solution fuel to a heterogeneous system using fuel rods in order to obtain criticality characteristics of fuel debris. The modification of the STACY was completed in December 2023. A series of performance inspections were conducted for the start of experimental operations. A new thermal power calibration is required for the performance inspections in order to operate at less than 200 W, which is the permitted thermal power. However, the thermal power measurement method and calibration data used in the former STACY is no longer available due to the modification of the modified STACY. We measured the thermal power of the STACY using the activation method that was improved to adapt to the measurement condition and calibrated the power meter system. Since the positions where activation foils could be installed were very limited, the thermal power was evaluated using numerical calculations supplemented by experimental data. Neutron flux data at the positions of the activation foil was measured by the activation method. Neutron distribution in the core was calculated by the Monte Carlo code MVP. A response function of the activation foil was calculated using the PHITS. The uncertainty of the thermal power measurement was conservatively estimated to be about 15%. Four operations were conducted for the thermal power measurement. The power meter was calibrated by using three operational data and tested with the one operational data. It was found that the indicated value of the meter adjusted by the STACY before the modification work would tend to overestimate the actual output by about 40%. In addition, the current calibration was able to calibrate the meter to within 3% accuracy.
Hamdani, A.; Soma, Shu; Abe, Satoshi; Shibamoto, Yasuteru
Progress in Nuclear Energy, 185, p.105771_1 - 105771_13, 2025/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Ogawa, Tatsuhiko
Annals of Nuclear Energy, 216, p.111256_1 - 111256_12, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)A novel robust method has been developed to simulate the performance of composite neutron sources composed of an alpha-emitting actinide and a light nucleus with low neutron separation energy. This method is based on the JENDL-5 cross-section data library and the Monte-Carlo radiation transport code PHITS. In contrast to previously devised methods, this approach can predict various quantities of the sources, such as actinide grain size dependence, absolute neutron emission intensity, energy spectra of neutrons and parasitic photons, neutron multiplicity, and time structure, with little approximation. The accurate calculation of stopping power of alpha rays in actinide grains and light elements, as well as the use of (
,n) reaction evaluated cross sections, which is one of the unique features of PHITS Ver.3.34 and its later versions, are the essences of the method. This method allows for the calculation of quantities important for practical applications, such as detection signal frequency, coincidence event rate, and the impact of parasitic gamma-rays.
Sugita, Yutaka; Ono, Hirokazu; Beese, S.*; Pan, P.*; Kim, M.*; Lee, C.*; Jove-Colon, C.*; Lopez, C. M.*; Liang, S.-Y.*
Geomechanics for Energy and the Environment, 42, p.100668_1 - 100668_21, 2025/06
Times Cited Count：1 Percentile：65.19(Energy & Fuels)The international cooperative project DECOVALEX 2023 focused on the Horonobe EBS experiment in the Task D, which was undertaken to study, using numerical analyses, the thermo-hydro-mechanical (or thermo-hydro) interactions in bentonite based engineered barriers. One full-scale in-situ experiment and four laboratory experiments, largely complementary, were selected for modelling. The Horonobe EBS experiment is a temperature-controlled non-isothermal experiment combined with artificial groundwater injection. The Horonobe EBS experiment consists of the heating and cooling phases. Six research teams performed the THM or TH (depended on research team approach) numerical analyses using a variety of computer codes, formulations and constitutive laws.
