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Kondo, Ryoichi; Nagaya, Yasunobu
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2023) (Internet), 10 Pages, 2023/08
A functional expansion tally (FET) method with numerical basis functions generated by singular value decomposition (SVD) is newly proposed. Traditionally, analytical functions were used for the FET calculations, e.g., Legendre polynomials for a one-dimensional distribution. However, the expansion terms could increase to reconstruct steep or complex distributions with these functions. A basis set that can well represent the target distribution with lower order expansion is desired to achieve high accuracy with the small computational resource. In the present study, a numerical basis set is generated from snapshot data by using SVD. This approach is based on the reduced order modeling (ROM). We applied ROM to the FET method in Monte Carlo calculations. The numerical result showed the applicability of the proposed method, on the other hand, some issues were revealed, e.g., discretization of the snapshot data.
Tada, Kenichi; Kondo, Ryoichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(6), p.624 - 631, 2023/06
Times Cited Count：2 Percentile：53.91(Nuclear Science & Technology)The sensitivity analysis and the uncertainty quantification have an important role in improving the evaluated nuclear data library. The current computational performance enables us to the sensitivity analysis and uncertainty quantification using the continuous energy Monte Carlo calculation code. The ACE file perturbation tool was developed for these calculations using modules of FRENDY. This tool perturbs the microscopic cross section, the number of neutrons per fission, and the fission spectrum. The uncertainty quantification using the random sampling method is also available if the user prepares the covariance matrix. The uncertainty of the k-effective using the perturbation tool was compared to the current sensitivity analysis codes SCALE/TSUNAMI and MCNP/KSEN. The comparison results indicated that the random sampling method using this tool accurately estimates the uncertainty of k-effective.
Kondo, Ryoichi
Robutsuri No Kenkyu (Internet), (75), 3 Pages, 2023/03
The author received the encouragement award from the reactor physics division of the atomic energy society of Japan. The research on the RSE (Resonance calculation using energy Spectrum Expansion) method, which is the subject of the award, is described for the bulletin of the reactor physics division.
Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Konno, Chikara; Kondo, Ryoichi; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*
Journal of Nuclear Science and Technology, 10 Pages, 2023/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Nuclear data processing code is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g., neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, consideration of the resonance upscattering, ACE file perturbation, and modification of ENDF-6 formatted file. FRENDY version 2 was released including these new functions. It generates GENDF and MATXS formatted neutron multi-group cross section files from an ACE formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.
Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi
Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.1493 - 1502, 2019/00
A perturbation capability of ACE formatted cross section files was developed using the modules of FRENDY. Uncertainty quantification using MCNP was carried out for the Godiva critical experiment by the RS method. We verified the results of the RS method by comparing with those obtained by the conventional sensitivity analyses. Moreover, uncertainty reduction using the bias factor method with the RS technique was applied to kinetic parameter, i.e., neutron generation time.
Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi
no journal, ,
The random sampling module for ACE format cross sections are developed using modules of FRENDY. This module perturbs the cross sections and other parameters in ACE format cross section library using covariance data. The GODIVA reactor is used and the calculation results of TSUNAMI-1D are compared to verify this module.
Kondo, Ryoichi; Nagaya, Yasunobu
no journal, ,
A functional expansion tally (FET) method expanded by numerical basis functions has been developed for Monte Carlo transport simulation. The numerical basis functions were extracted from various flux distributions by singular value decomposition, to expand the target flux distribution with low order bases. In this work, multi-group Monte Carlo calculations were carried out for one-dimensional geometry. The accuracy of the spatial flux distributions obtained by the proposed method was confirmed in comparison with the traditional discrete cell tally method and the FET method expanded by Legendre polynomials.
Kondo, Ryoichi; Nagaya, Yasunobu
no journal, ,
A functional expansion tally (FET) method expanded by numerical basis functions is under development for Monte Carlo transport simulation. In this work, a multi-group Monte Carlo calculation was performed to obtain the flux distribution with the FET method using numerical basis functions for one-dimensional whole core geometry. The numerical basis functions were generated by singular value decomposition of fluxes, which were calculated by a deterministic method in unit assembly with various calculation conditions. The whole core flux distribution was calculated by expanding the flux distribution of each assembly with the numerical basis functions. The accuracy of the proposed method was confirmed in comparison with the discrete tally method and the conventional Legendre polynomials based FET method.
Tada, Kenichi; Kondo, Ryoichi; Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
no journal, ,
JAEA has developed a Python-based multi-physics platform JAMPAN. This platform has an HDF5 formatted JAMPAN data container. It connects calculation codes via this data container. The utilization of this container eliminates the dependence on the calculation code and enables us to easily exchange the coupling code. The first target of JAMPAN is a coupling of neutronics and thermal hydraulics codes to provide reference results of core analysis codes. The coupling of the other codes such as a fuel performance analysis code FEMAXI is future work. This presentation shows the overview of the JAMPAN platform.
8 single bundle assembly of BWRKamiya, Tomohiro; Ono, Ayako; Nagatake, Taku; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki
no journal, ,
JAEA aims to obtain reference solutions for reactor design codes by coupling the Monte Carlo code MVP and the multiphase and multi-component detailed thermal-hydraulic analysis code JUPITER on the multiphysics platform JAMPAN (JAEA Advanced Multi-Physics Analysis platform for Nuclear systems). For BWR, the thermal-hydraulic analysis code is required to consider boiling around fuel rods. Therefore, a thermal-hydraulic simulation of an 88 STEP-II single fuel assembly system was performed considering boiling using the temperature recovery method.
Tada, Kenichi; Kondo, Ryoichi; Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
no journal, ,
JAEA has developed the multi-physics platform JAMPAN. In the previous presentation, we demonstrated a BWR single fuel assembly calculation by the coupling calculation of the continuous energy Monte Carlo calculation code MVP and the subchannel analysis code NASCA. The final goal of the MVP/NASCA coupling calculation is the whole core analysis. To achieve this, we implemented the flow rate calibration function in JAMPAN for the MVP/NASCA coupling calculation of the BWR multi-fuel assembly geometry.
Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki
no journal, ,
JAEA has developed a platform JAMPAN for multi-physics simulations, has improved a neutronics analysis code, and has improved and validated thermal-hydraulics analysis codes to improve the design and the safety of light water reactors. The objective is implementing the coupling modules between the neutronics code MVP and the thermal-hydraulics code JUPITER, and verifying the modules. A fuel bundle geometry under a normal operation condition of a BWR was used for the neutronics and thermal-hydraulics coupling simulation to verify the modules. In this presentation, we will explain how to send and receive data between MVP and JUPITER through JAMPAN and show the results of the neutronics/thermal-hydraulics coupling simulations using MVP and JUPITER.
