Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Mori, Takero; Yoshimura, Kazuo; Yoshikawa, Ryuji; Kikuchi, Norihiro; Matsushita, Kentaro; Mochinaga, Shota; Ezure, Toshiki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR26) (Internet), 9 Pages, 2026/05
This paper presents the outline of the evaluation support program on the evaluations of the thermal-hydraulics related issues in the conceptual design study of the sodium-cooled fast reactor for demonstration in Japan, utilizing the technical basis on the simulations and evaluations configured in the integrated evaluation system named ARKADIA that has been developed with experiences through the development of the SFRs and related R&Ds in Japan Atomic Energy Agency. Moreover, as the extension of the function of the ARKADIA, the possible framework for SFR lifecycle development introducing the Digital-Twin for the MBD with ARKADIA is presented with an example of a whole plant coupled analysis model developed in ARKADIA.
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.
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.
Kikuchi, Norihiro; Yoshikawa, Ryuji; Tanaka, Masaaki
Proceedings of 32nd International Conference on Nuclear Engineering, Vol.15 (Internet), p.647 - 659, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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.
Yoshikawa, Ryuji; Kikuchi, Norihiro; Tanaka, Masaaki
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
In the study of safety enhancements on advanced sodium-cooled fast reactor, it has been essential to evaluate the influence of buoyancy on pressure drop in a fuel assembly at mixed convection condition during natural circulation under the decay heat removal operation. In this study, the numerical simulations of the 19-rod and 91-rod bundle water experiments at low flow rate conditions were performed for the validation of a thermal-hydraulic analysis code named SPIRAL with the hybrid turbulence model. The influence of buoyancy on the velocity and temperature distributions was analyzed, and the applicability of the hybrid turbulence model to the pressure drop evaluation was investigated by comparison with the experimental friction factors.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Nuclear Technology, 210(5), p.814 - 835, 2024/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.
Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki
Journal of Nuclear Engineering and Radiation Science, 9(3), p.031401_1 - 031401_11, 2023/07
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) investigated in JAEA, the use of a specific fuel assembly with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Since the fuel rods have an asymmetric layout by the inner duct, the validity confirmation of the numerical results of an in-house subchannel analysis code named ASFRE was required. In this paper, therefore, the code-to-code comparisons was applied with numerical results of ASFRE and those of an in-house CFD code named SPIRAL. The applicability of ASFRE was indicated through the confirmation of the consistency of specific temperature distributions.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Removal of core decay heat by utilizing natural circulation is expected as a significant measure to enhance the safety of sodium-cooled fast reactors (SFRs). Accurate evaluation of the temperature distribution in the fuel assembly (FA) at the low Re regime in natural circulation operation is demanded. A detailed thermal-hydraulics analysis code named SPIRAL has been developed to clarify thermal-hydraulic phenomena in the FA at various operation conditions. In this study, SPIRAL with the hybrid turbulence model was applied to analyze a large-scale fuel assembly experiment of a 91-pin bundle for two cases at the mixed and the natural convection conditions respectively in low Re regime with heat transfer from outside of the FA. The applicability of the SPIRAL to the thermal-hydraulics evaluation of FA at mixed and natural convection conditions was confirmed by the comparisons of temperatures predicted by SPIRAL with those measured in the experiment.
nard convection using new developed mist modelsOhira, Hiroaki*; Tanaka, Masaaki; Yoshikawa, Ryuji; Ezure, Toshiki
Annals of Nuclear Energy, 172, p.109075_1 - 109075_10, 2022/07
Times Cited Count：1 Percentile：9.19(Nuclear Science & Technology)In order to evaluate the mist behavior in the cover gas region of Sodium-cooled Fast Reactors (SFRs) in good accuracy, turbulent model for Rayleigh-Bnard convection (RBC) was selected, and the Reynolds-averaged number density and momentum equations for mist behavior were developed and incorporated into the OpenFOAM code. In the first stage, the RBC in a simple parallel channel was calculated using Favre-averaged k-
SST model. The average temperature and flow characteristics agreed well with results from DNS, LES, and experiments. Then the basic heat transfer experiment simulating the cover gas region of SFRs was calculated using this turbulent model and new mist models. The calculated average temperature distribution in the height direction and the mist mass concentration agreed well with the experimental results. We developed a method that could simulate the mist behavior in turbulent RBC environments and the cover gas region of SFRs with high accuracy.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in JAEA, the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. For the FAIDUS, confirmation of validity of the numerical results by a subchannel analysis code named ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. Therefore, the code-to-code comparisons with numerical results of ASFRE and those of a CFD code named SPIRAL was conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10
A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Journal of Nuclear Engineering and Radiation Science, 5(2), p.021001_1 - 021001_12, 2019/04
In the design study of an advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been considered as one of the measures to enhance safety of the reactor during the core disruptive accident. In this study, thermal-hydraulics in FAIDUS was investigated with the in-house subchannel analysis code named ASFRE. Before the application to FAIDUS, applicability of ASFRE to FAs was confirmed through the numerical simulations for the experiments of simulated FA. Through the comparisons between the numerical results of thermal-hydraulic analyses of FAIDUS and a typical FA without the inner duct, it was indicated that significant asymmetric temperature distribution would not occur in FAIDUS at both high and low flow rate conditions.
Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08
A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07
In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.
Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki; Imai, Yasutomo*
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10
A sodium-water coupled thermal-hydraulics simulation code TSG has been developed for numerical estimation of three-dimensional thermal-hydraulic phenomena in the straight-tube steam generator. The water analysis module was developed by using the parallel channel model of heat transfer tubes, and the sodium analysis module was developed by using porous body approach. As the first step of validation, simulation results by TSG were compared with the measured data of 1MWt SG experiments under steady state conditions. Through the numerical simulation, the coupled simulation method used in TSG was validated and applicability of TSG to simulate thermal-hydraulics of the straight tube SG in the steady state was confirmed.
Yoshikawa, Ryuji; Ohshima, Hiroyuki; Tanaka, Masaaki; Imai, Yasutomo*
JAEA-Data/Code 2014-034, 84 Pages, 2015/03
JAEA-Data-Code-2014-034.pdf:2.35MB
TSG (Three-dimensional Thermal-hydraulics Analysis Code for Steam Generators) is being developed for analyses of thermal hydraulics in double wall straight tube steam generator of Fast Breeder Reactor. TSG code is a thermal hydraulics simulation system which couples sodium side three dimensional simulation with water side multi-channel simulation. The three dimensional flow field in the sodium side is simulated by a commercial code FLUENT with porous media model. The multi-channel two-phase flow is simulated by an in-house module with drift-flux model. The sodium side simulation is coupled with the water side simulation by the transmission of heat transfer rate through the heat transfer tube. This report presents a description of the computational models, input and output as the user's manual of TSG water side module.
