A Study on the applicability of uncertainty quantification and sensitivity analysis in validation process for fast reactor plant dynamics analysis code

Hamase, Erina; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki

Annals of Nuclear Energy, 236, p.112358_1 - 112358_13, 2026/10

https://doi.org/10.1016/j.anucene.2026.112358

To 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.

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.

Applicability of uncertainty quantification and sensitivity analysis for validation of fast reactor plant dynamics analysis code

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.

Validation study on SFR core bowing codes using Joyo ex-core experiment data; Single duct bowing benchmark

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

https://doi.org/10.15669/pnst.8.160

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.

Core deformation reactivity with neutronics-thermal hydraulics-structural mechanics coupled analysis for FFTF LOFWOS Test #13

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.

Development of ARKADIA for the innovation of advanced nuclear reactor design process (Development of the design optimization support tool, ARKADIA-Design)

Tanaka, Masaaki; Doda, Norihiro; Hamase, Erina; Kuwagaki, Kazuki; Mori, Takero; Okajima, Satoshi; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Hashidate, Ryuta; et al.

Nihon Kikai Gakkai Rombunshu (Internet), 91(943), p.24-00229_1 - 24-00229_12, 2025/03

https://doi.org/10.1299/transjsme.24-00229

To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) has been developed. In this paper, focusing on the ARKADIA-Design, achievements in the development of optimization processes in the fields of the core design, the plant structure design, and the maintenance schedule planning, as major function of ARKADIA-Design, and numerical analysis methods including coupled analysis to be used for the detailed analysis to confirm the plant performance after optimization are introduced at this point in time.

Development of core design optimization process; Feasibility study of multivariable optimization via integrated sequential analyses of neutronics, thermal-hydraulics, and fuel integrity evaluation

Kuwagaki, Kazuki; Hamase, Erina; Yokoyama, Kenji; Doda, Norihiro; Tanaka, Masaaki

Annals of Nuclear Energy, 225, p.111754_1 - 111754_10, 2025/01

https://doi.org/10.1016/j.anucene.2025.111754

Development of VVUQ method for ensuring credibility of plant dynamics analysis results based on statistical approach

Hamase, Erina; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 10 Pages, 2024/11

A plant dynamics analysis code, Super-COPD, is being developed for the design and safety evaluation of sodium-cooled fast reactors. Verification, validation, and uncertainty quantification (VVUQ) are required to ensure the reliability of its analysis results. In this study, to develop the VVUQ method, the uncertainty propagation analysis of input parameters was performed for the loss of flow without scram test in the FFTF, and the process of validation was investigated. In addition, the method of sensitivity analysis was investigated. As a result, the uncertainty of the analysis results was quantified, the applicability of the statistical method was confirmed. The sensitivity analysis using the Sobol' method identified the models that needs to be prioritized for improvement.

Validation study on SFR core bowing codes using Joyo ex-core experiment data; Multiple duct bowing benchmark

Wozniak, N.*; Shemon, E.*; Feng, B.*; Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; et al.

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies.

Development of reactor vessel thermal-hydraulic analysis method in natural circulation conditions; Applicability investigation for transient analysis

Hamase, Erina; Miyake, Yasuhiro*; Imai, Yasutomo*; Doda, Norihiro; Ono, Ayako; Tanaka, Masaaki

Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09

In a design study of sodium-cooled fast reactors, we have developed the practical reactor vessel thermal-hydraulic analysis method (RV-CFD) that had a low computational cost about the thermal-hydraulics in the core to evaluate the core-plenum interactions occurred in the natural circulation decay heat removal during the dipped-type direct heat exchanger operation. In this study, the non-equilibrium thermal model which considered the thermal inertia of fuel pins was developed and incorporated into the core of RV-CFD. Through the transient analysis simulating the power reduction due to reactor scram, the applicability of RV-CFD to the transient analysis was confirmed.