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Yada, Hiroki; Takaya, Shigeru; Machida, Hideo*
Mechanical Engineering Journal (Internet), 13(2), p.24-00457_1 - 24-00457_14, 2026/04
Yoshikawa, Ryuji; Imai, Yasutomo*; Tanaka, Masaaki
JAEA-Research 2025-015, 100 Pages, 2026/03
JAEA-Research-2025-015.pdf:6.94MB
TSG (Three-dimensional Thermal-hydraulics Analysis Code for Steam Generators) has been developed for the numerical simulation of thermal hydraulics in double wall straight tube steam generator (SG) of Sodium-cooled Fast Reactor (SFR) by the Japan Atomic Energy Agency (JAEA). TSG is a thermal hydraulics simulation system for double wall straight tube SG which couples the sodium side three-dimensional simulation with water side multi-channel simulation. The three-dimensional flow field in sodium side is simulated by the CFD code FLUENT with porous media model. The multi-channel two-phase flow in water side is simulated by in-house code with drift-flux model. The sodium side simulation is coupled with water side simulation by the transmission of heat transfer rate through the heat transfer tube, therefore the overall thermal hydraulics in SG can be evaluated transiently. This report presents the sodium-water coupled simulation models of TSG, and the simulation results of fundamental validation of TSG with the steady state results of 1MWt SG tests. Next, the evaluation results of temperature deviation at the heat transfer tube plugging conditions in a straight tube SG of a commercial reactor, and the evaluation results of three-dimensional temperature distribution and structural integrity at the heat transfer tube plugging condition for the large-sized SG including the inlet and outlet plenums are described. In addition, the applicability of TSG to the flow stability analyses for 1MWt SG instability tests is presented in the appendix.
Okamoto, Naritoshi; Komeno, Akira; Seya, Atsumasa; Inaba, Hideki*; Terakado, Shinichi*; Higuchi, Masashi*
JAEA-Data/Code 2025-022, 497 Pages, 2026/03
JAEA-Data-Code-2025-022.pdf:18.06MB
The Plutonium Fuel Third Development Laboratory of the Nuclear Fuel Cycle Engineering Laboratories has applied for a change of use permit (hereinafter referred to as "license") for plutonium fuel facilities. For the criticality safety design of gloveboxes and equipment/instruments handling mixed oxide (MOX), various criticality calculation codes are used. The most recent employs the 3D Monte Carlo calculation code KENO-V.a embedded in the SCALE 4.4 code system, along with the 27-group ENDF/B-IV neutron cross-section library. SCALE 4.4 was released by the Oak Ridge National Laboratory (ORNL) in the US in 1998, and has now been in use for 27 years. ORNL has continuously improved its functionality, with SCALE 6.3.2 released in 2024. When designing and constructing new MOX fuel facilities, it is desirable to obtain a license using criticality calculation codes based on the latest knowledge. However, it is necessary to verify that these codes have sufficient reliability. Therefore, in 2018, benchmark calculations were performed using the 252-group ENDF/B-VII.1 neutron cross-section library (v7-252n) for two versions of the criticality calculation sequences KENO-V.a and KENO-VI from SCALE 6.2.3, based on past criticality experimental setups. The estimated critical-limiting multiplication factor was calculated. The results indicate that these codes can be used with sufficient confidence for criticality safety design of MOX fuel facilities.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
JAEA-Data/Code 2025-018, 96 Pages, 2026/03
JAEA-Data-Code-2025-018.pdf:5.54MB
In the Japan Atomic Energy Agency, a detailed thermal-hydraulic analysis code named SPIRAL based on the finite element method (FEM) is being developed to evaluate the detailed thermal-hydraulic properties of fuel assemblies (FAs) in sodium-cooled fast reactors (FBRs). Because the quality of the computational grid (elements) used in the calculations has a significant impact on the prediction accuracy, the allocation of high-quality elements in the wire-spacer-type FA pin bundle region is an important issue for numerical analysis. Although a commercial mesh generation program (mesher) with CAD data of FA's geometric shape can be considered as one measure, it is an extremely complicated task to perform element division of complex FA region. Therefore, to efficiently allocate high-quality elements, we developed a mesher that automatically performs element division in the FA region using the FA's geometric shape (design information) and meshing parameters as input conditions. This report describes the details of the mesher's various meshing models and their usage. To regularly allocate the computational grid for the complex FA region, the mesher first divides the region into multiple blocks using a multi-block method, then generates boundary-fitted curvilinear coordinate grids for each block region, and finally integrates them into a single FA mesh system. In addition, a combination of hexahedral elements and prism-shaped elements is arranged to maintain element continuity between adjacent block regions. Element division for both the normal FAs surrounded by a hexagonal cross-section tube and the irregular FAs, inside which a duct is installed to promote the discharge of molten fuel, is possible. The development of this mesher has made it possible to accurately and efficiently perform element division of complex FA region on various conditions.
/k
- modelKikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
JAEA-Data/Code 2025-017, 133 Pages, 2026/03
JAEA-Data-Code-2025-017.pdf:3.9MB
In a core design of sodium-cooled fast reactors (SFRs), it is necessary to confirm the integrity of fuel assemblies (FAs) in the core over a wide range of operating conditions. To evaluate the velocity and temperature distributions within the FAs in detail, we have been developing a detailed FA thermal-hydraulic analysis code named SPIRAL. In our previous works, we implemented numerical methods for fluid mechanics at isothermal conditions and turbulence models. Subsequently, we implemented turbulent heat transfer models for the evaluation of temperature distribution within the FAs, and validated them through experimental analyses mainly under high flow rate conditions. The thermal-hydraulics within the FAs varies depending on the operating conditions. Furthermore, the local Reynolds (Re) number within the FAs varies widely due to the influence of wire spacers spirally wound around the fuel rod. For this reason, it has been shown that standard and low Re number k-/k- models have difficulty reproducing the thermal-hydraulics in the laminar-turbulent transition region. Therefore, to reproduce the thermal-hydraulics over a wide Re number range, we developed a hybrid k-/k- model that combines the standard k-/k- model with the advantages of the low Re number k-/k- model. This paper describes the governing equations, constitutive equations derived from various turbulence models, their formularizations by the finite element method, their numerical treatment, and the treatment of boundary conditions. We also report the results of analyses conducted to validate the hybrid k-/k- model for predicting pressure drop and temperature distribution.
Mikami, Nao; Aizawa, Kosuke; Kurihara, Akikazu; Ueki, Yoshitaka*
AI Thermal Fluids (Internet), 5, p.100029_1 - 100029_15, 2026/03
Yamashita, Hayato; Toyota, Kodai; Onizawa, Takashi; Yamamoto, Kenji*; Kubo, Koji*
Nihon Kikai Gakkai Rombunshu (Internet), 92(955), p.25-00176_1 - 25-00176_13, 2026/03
It is planned that Mod.9Cr-1Mo steel will be used as the structural material for the steam generator of the demonstration fast reactor. Creep strength of welded joints of Mod.9Cr-1Mo steel is lower than that of base metal at high temperature and long time. In addition, the creep strength of welded joints of Mod.9Cr-1Mo steel with repair welding is lower than that of without repair welding. In this study, the effects of the location and number of repair welds performed on the creep strength of welded joints of Mod.9Cr-1Mo steel were investigated, and a repair welding method was developed for the construction of the demonstration fast reactor. It was clarified that, regardless of the location and number of repair welds performed, the degradation in creep strength could be greatly reduced by performing the repair welding before post weld heat treatment (PWHT). Furthermore, it was found that the superposition of thermal history and the formation of a heat affected zone (HAZ) on the weld metal promote the coarsening of ferrite grains during creep and cause a slight reduction in creep strength. All repair welding was performed prior to PWHT, and a repair welding method was also developed to minimize thermal history superposition and weld metal HAZ formation.
Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Fujisaki, Tatsuya*; Murakami, Satoshi*
Annals of Nuclear Energy, 226, p.111896_1 - 111896_11, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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：0 Percentile：0.00(Nuclear Science & Technology)
Th as a long-life 
Ac generator using the experimental fast reactor JoyoSasaki, Yuto; Maeda, Shigetaka; Fukasawa, Tetsuo*; Takaki, Naoyuki*
Journal of Nuclear Science and Technology, 63(2), p.154 - 165, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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.
Mori, Tetsuya; Oki, Shigeo
Nuclear Technology, 212(2), p.490 - 509, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study investigates the characteristics of the Doppler coefficient and sodium void reactivity of a burning fast reactor core concept, which was constructed in a previous study. This concept allows for multiple recyclings of plutonium and minor actinides (transuraniums (TRU)). TRU degradation due to multiple recycling deteriorates the reactivity coefficients through indirect effects, such as by hardening the neutron spectrum and steepening the energy gradient of neutron importance. Using silicon carbide (SiC) structural material improves the reactivity coefficient by causing an opposite indirect effect of TRU degradation. This improvement results not only from neutron spectrum softening due to the neutron moderation effect from 
C but also from the neutron leakage effect resulting from the low structural material density. The disadvantage of increased calculation uncertainty by using SiC structural material can be practically ignored. Furthermore, the burning core has Doppler coefficient enhancement characteristics by the moderated neutron reflection effect from outside the core. This characteristic has the potential to provide a new measure for reactivity coefficient deterioration due to TRU degradation. The reactivity coefficient characteristics clarified in this study can provide valuable knowledge for future detailed designs and design improvements of a TRU burning core.
Kam, D. H.*; Grabaskas, D.*; Okano, Yasushi; Uchibori, Akihiro; Starkus, T.*
Nuclear Technology, 212(2), p.347 - 364, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Nakahara, Masaumi; Yamagata, Ryohei*; Ishii, Yasuyuki*; Koka, Masashi*
QST-M-56; QST Takasaki Annual Report 2024, P. 53, 2026/02
In the minor actinides recovery process from high-level liquid waste, analysis for complex structure in organic solvent has been studied to strip minor actinides from a loaded solvent efficiently. In this study, Eu was used as a simulated material of minor actinides, and organic solvents containing Eu complexes were prepared. The composition of Eu complexes in organic solvents was confirmed by particle induced X-ray emission. In addition, the ion beam induced luminescence was measured, and the basic data for the complex structure were obtained.
-PuO
-PuO systemVinograd, V. L.*; Vauchy, R.
Journal of Nuclear Materials, 619, p.156244_1 - 156244_16, 2026/01
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)Thermodynamic properties of (U
Pu
)O
fluorite (FCC) and (UPu)O bixbyite (BCC) in the UO-PuO-PuO system are assessed by considering phase equilibrium constraints and data on the variation of oxygen to metal ratio (O/M) as a function of the chemical potential of O. Thermodynamically, both BCC and FCC are described as ordered solid solutions allowing for a decrease in the configurational entropy of their oxygen/vacancy distributions at the specific values of 
= -0.5 and = -0.375 (O/M = 1.5 and O/M = 1.625). With this approach, fluorite/bixbyite equilibria in PuO-PuO and in UO-PuO-PuO are reproduced well with little effort. Moreover, we show that a large manifold of experimental data on the UO-PuO-PuO system is consistent with the assumption that Pu/(Pu+U) ratios in individual phases remain equal to the total Pu/(Pu+U) ratio in the system, i.e., no inter-phase U/Pu-partitioning occurs.
Maruyama, Shuhei; Yamamoto, Akio*; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 63(1), p.31 - 44, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Ishida, Shinya; Tagami, Hirotaka*; Tobita, Yoshiharu*; Kawada, Kenichi; Morita, Koji*
Nihon Genshiryoku Gakkai-Shi ATOMO
, 68(1), p.16 - 20, 2026/01
no abstracts in English
Miyahara, Shinya*; Koie, Ryusuke*; Uno, Masayoshi*; Kawaguchi, Munemichi*; Sato, Rika; Seino, Hiroshi
Nuclear Engineering and Design, 446(Part A), p.114523_1 - 114523_14, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Kikuchi, Norihiro; Yoshikawa, Ryuji; Tanaka, Masaaki
Proceedings of 32nd International Conference on Nuclear Engineering, Vol.15 (Internet), p.647 - 659, 2026/01
An in-house subchannel analysis code called ASFRE have been developed to evaluate fuel assembly (FA) thermal-hydraulics in sodium-cooled fast reactors (SFRs). In this study, models to solve the important phenomena in the FA and necessary experiments for validation were listed systematically in order to assess the reliability of the codes, through developing an importance ranking table for the phenomena and a validation matrix according to the guide-line for the verification and validation (V&V). The ranking table was developed to decide the priority for validation. In addition, a validation matrix of experimental data and numerical models in the codes for the high priority phenomena in the ranking table were developed to confirm the sufficiency of the validation process.
Pu
Am
O
Vauchy, R.; Horii, Yuta; Hirooka, Shun; Akashi, Masatoshi; Sunaoshi, Takeo*; Nakamichi, Shinya; Saito, Kosuke
Proceedings of 34th Nuclear Energy for New Europe (NENE2025) , p.232 - 238, 2026/01
Ueki, Yoshitaka*; Hirako, Itsuki*; Tezuka, Kosuke*; Aizawa, Kosuke; Ara, Kuniaki*
AI Thermal Fluids (Internet), 4, p.100021_1 - 100021_12, 2025/12
