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Hamase, Erina; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki
Annals of Nuclear Energy, 236, p.112358_1 - 112358_13, 2026/10
Times Cited Count：0To ensure the reliability of analysis results from the plant dynamics analysis code Super-COPD, a validation process comprising forward uncertainty quantification (Forward UQ) and sensitivity analysis (SA) using the Sobol method was developed. Uncertainty propagation analysis of input parameters was performed for the loss of flow without scram test in the FFTF and demonstrated that encompassing test results can serve as one measure validation criterion. Furthermore, SA identified dominant input parameters affecting uncertainty and provided effective targets for reducing uncertainty. This study confirms that Forward UQ and SA using the Sobol method are applicable for the validation process.
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.
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.
Endo, Tatsuro*; Araki, Yasufumi; Seki, Munetoshi*; Tabata, Hitoshi*; Tanaka, Masaaki*; Oya, Shinobu*
Applied Physics Letters, 128(2), p.022405_1 - 022405_6, 2026/01
Times Cited Count：0 Percentile：0.00(Physics, Applied)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.
Fukui, Makoto; Chizuwa, Shingo*; Kikuchi, Norihiro; Tanaka, Masaaki; Hashimoto, Makoto
JAEA-Review 2025-045, 42 Pages, 2025/12
JAEA-Review-2025-045.pdf:2.95MB
Hot laboratory (HL) at Oarai Nuclear Engineering Institute is a facility that conducts post-irradiation testing of fuel samples and reactor materials in hot cells. A set of local sampling system (LSS) is installed as a radiation control equipment to monitor the concentration of radioactive materials in the air in work environment. The LSS of HL equipped 23 sampling points, which are called as local sampling ends (LSE). It was recognized that air sampling had not operated at some of the LSE, and the concentration of radioactive materials in the air was not measured as prescribed. In this report, we evaluated the effect of partial termination of the LSS and the resulting increase in sampling intervals on the control of radioactive material concentrations in the air using airflow analysis assuming the diffusion of radioactive materials from hot cells in the controlled area of HL. The Service Area of the HL, where 10 LSEs were set in a wide area, was selected as an evaluation area. Airflow analysis including the diffusion of virtual contaminant particles was conducted on the evaluation area. Diffusion of virtual contaminants from hot cells and sampling of virtual contaminants at LSEs are simulated in the case of LSS in fully working and LSS with termination of 4 LSEs. The evaluation results showed that the effect of the partial termination of LSS and the resulting increase in sampling intervals on the control of the concentration of radioactive materials in the air are small.
Renaudi
e de Vaux, S.*; Li, S.*; Marrel, A.*; Ezure, Toshiki; Tanaka, Masaaki
Nuclear Engineering and Design, 444, p.114382_1 - 114381_15, 2025/12
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In this study, experimental results on sodium natural circulation in PLANDTL-2 under eight transient scenarios were reported. The experimental parameters were the initial state of the primary circuit, the shutdown inertia of the secondary circuit, and the operating mode of the Decay Heat Removal system (DHRS). As the results, it was clarified that flow reversals can be observed in the primary circuit under rapid cooling cases by DHRS linked to an enhancement of thermal stratification in the plenum. A novel method was introduced to estimate the delay between DHR system activation and increase in stratification. In addition, some guidelines for future investigations were given in order to avoid the presently described instabilities.
Tanaka, Masaaki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 67(10), p.588 - 592, 2025/10
The numerical analysis and evaluation technology base, including thermal hydraulic CFD technology is being developed in the integrated evaluation system named "ARKADIA" at the Japan Atomic Energy Agency. Regarding the deployment of thermal hydraulic CFD technology for the development of sodium-cooled fast reactors (SFR), application strategy of the technology base in ARKADIA to the conceptual design process of the demonstration SFR was briefly introduced, touching on the experiences of the application study to the new regulatory standards for "Joyo". The development of knowledge (KMS) - simulation (VLS) linkage function to concrete capitalization and effectively utilization of the experiences (knowledge) of the numerical analyses in JAEA was also introduced. And furthermore, the extension of the whole plant analysis method being developed in ARKADIA to the digital twin technology in the SFR development was mentioned.
Alzahrani, H.*; Matsushita, Kentaro; Sakai, Takaaki*; Ezure, Toshiki; Tanaka, Masaaki
Nuclear Technology, 211(10), p.2446 - 2458, 2025/10
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Development of evaluation method for cover gas entrainment by vortices generated at free surface in upper plenum of sodium-cooled fast reactor is required, and an evaluation method by predicting vortices from flow velocity distribution obtained by CFD analysis is developed. In this study, Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis. Initial mesh was refined with two indexes: the first index (Index-1) is when the second invariant of velocity gradient tensor, Q, is negative and the second one (Index-2) is pressure gradient index added to Index-1. As a result of applying AMR method to unsteady vortices system with a flat plate and performing transient analyses with refined meshes, the result of pressure distribution and velocity around the flat plate in mesh using Index-2 was similar to the result of all refined mesh. It was also confirmed that vortices generation and growth was better simulated by refining meshes around separation area.
Hamase, Erina; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki
Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09
To investigate the applicability of uncertainty quantification (UQ) and sensitivity analysis (SA) methods for validating a fast reactor plant dynamics analysis code, forward UQ and SA employing Sobol' method were performed for FFTF LOFWOS test No.13. The result demonstrated that validity can be judged if the test results fall within the quantified uncertainty range, and that the dominant input parameters influencing uncertainty can be quantitatively evaluated, enabling prioritization of parameters for uncertainty reduction. This confirms the applicability of forward UQ and SA employing Sobol' method.
Matsushita, Kentaro; Ezure, Toshiki; Fujisaki, Tatsuya*; Nakamine, Yoshiaki*; Imai, Yasutomo*; Tanaka, Masaaki
Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09
In the design of sodium-cooled fast reactors (SFRs), it is important to evaluate the transition behavior of non-condensable gas entrained into the primary coolant system due to cover gas entrainment and dissolution. In this study, trajectories of non-condensable gas bubbles in the cold plenum of the pool-type SFR evaluated by three-dimensional CFD analyses applying Discrete Phase Model. As the result of sensitivity analyses regarding bubble radius flowing into the cold plenum, it was clarified that the release rate of bubbles showed an increase according to the increase of bubble radius and an asymptotic increasing behavior in the large bubble radius cases.
Ito, Kei*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Odaira, Naoya*; Ito, Daisuke*; Saito, Yasushi*
Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09
The estimation of entrained gas flow rate by a bathtub vortex is important in terms of a possibility to causes the performance degradation when the entrained bubbles are mixed into fluid machineries, e.g. pumps. In this study, to confirm the applicability of a model based on circulating annular flow model proposed by the authors, entrained gas flow rate is evaluated using the liquid velocity distribution around free surface dent of vortex (gas core), obtained by CFD data. As a result, it was indicated that it would be possible to evaluate the gas entrainment flow rate by setting an appropriate evaluation region.
Kato, Shinya; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Endo, Tomohiro*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
During a reactor power increase in ULOF and UTOP events in sodium-cooled fast reactors, core deformation due to thermal expansion of core elements is expected to cause a negative feed-back effect to suppress this power increase. An analytical evaluation method of core deformation reactivity for design is being developed in JAEA. However, the neutronics calculation module uses several approximations. This study aims to develop the detailed evaluation method as a reference neutron transport calculation code for confirmation of the validity of the calculated core deformation reactivity. Here, the two-dimensional finite volume method (FVM) code based on simplified P3 (SP3) approximation with unstructured mesh have been developed as the first step of the development. This paper describes the calculation theory of the FVM code, the procedure of introducing SP3 approximation into the code and the verification results of the functions developed.
Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Wozniak, N.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ogata, Takanari*; Shemon, E.*; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.160 - 164, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using a single duct from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement caused by thermal bowing as well as the contact load on the duct pad.
Wozniak, N.*; Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Shemon, E.*; Feng, B.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.165 - 170, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies in multiple duct configuration.
Kobayashi, Jun; Tanaka, Masaaki; Hamase, Erina; Ezure, Toshiki
JAEA-Data/Code 2025-009, 74 Pages, 2025/08
JAEA-Data-Code-2025-009.pdf:4.7MB
In a sodium-cooled fast reactor, a diversified auxiliary cooling system to remove decay heat from the core is required to enhance its safety. The decay heat removal systems (DHRSs) include a direct reactor auxiliary cooling system (DRACS) with a heat exchanger in the upper plenum (UP) of the reactor vessel (RV), a primary reactor auxiliary cooling system (PRACS) with a heat exchanger in the primary heat transport system (PHTS), an intermediate reactor auxiliary cooling system (IRACS) with a heat exchanger in the secondary heat transport system (SHTS), a heat removal system which employs a steam generator, and a reactor vessel auxiliary cooling system (RVACS) that effects cooling from outside the RV. In the operation of the DRACS with a dipped-type direct heat exchanger (D-DHX) installed in the UP of the RV (UP-RV), the thermal interaction, called core-plenum interaction (CPI), regarding the thermal-hydraulic phenomena in the UP-RV and the core is observed. The CPI includes the penetration flow of the sodium at a low temperature from the D-DHX into the core assemblies, the flow in the gap between assemblies, and the radial heat transfer through sodium in the gap. On the other hand, in the operation of the PRACS or IRACS, where the flowrate in the PHTS is maintained, the core coolability is affected by plant operating conditions. Two transient tests conducted at the PLANDTL-DHX sodium test facility in Japan Atomic Energy Agency were provided to develop an appropriate numerical analysis model for prediction of transient thermal-hydraulics in the DHRSs, the core, and the PHTS. In this document, the geometry information of the RV and the PHTS, including the heat exchangers for the DHRS, and the measured flowrate and temperature transients at each inlet of the intermediate heat exchanger (IHX) on the SHTS side and DHRS were specified as the boundary conditions for the benchmark analyses.
Hamase, Erina; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Imai, Yasutomo*
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 14 Pages, 2025/08
We have developed the reactor vessel thermal-hydraulic analysis model (RV-CFD) with the subchannel CFD (SC) model for assembly with a low computational cost to evaluate the core-plenum interactions occurred in the natural circulation decay heat removal during the dipped-type direct heat exchanger operation in sodium-cooled fast reactor. In this study, the non-equilibrium thermal model which can consider the heat capacity and thermal load of fuel pins was developed in the SC model. Through the transient analysis simulating the power reduction due to reactor scram using the RV-CFD, the applicability of RV-CFD to the transient analysis was confirmed.
Doda, Norihiro; Kato, Shinya; Uwaba, Tomoyuki; Tanaka, Masaaki; Nakamine, Yoshiaki*; Igawa, Kenichi*; Iida, Masaki*
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 14 Pages, 2025/08
Accurate evaluation of reactivity feedback due to core deformation during power increases in sodium-cooled fast reactors requires comprehensive modeling of the interactions among neutronics, thermal-hydraulics, and core mechanics. To accurately consider these interactions, JAEA has developed an evaluation method that combines multiple analysis codes that model these phenomena in detail. In this study, the evaluation method was applied to the core analysis of the FFTF LOFWOS Test #13, and the analysis results of net reactivity were compared with the test results. The sensitivity analysis results of the core structural design parameters showed that the core bowing behavior has a significant effect on the temporal variation of net reactivity.
