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Aizawa, Naoto*; Watanabe, Tomoaki; Chiba, Go*; Tada, Kenichi; Fujita, Tatsuya*; Yamamoto, Akio*
Nuclear Engineering and Technology, 58(5), p.104176_1 - 104176_16, 2026/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Kamiya, Tomohiro; Kondo, Ryoichi; Fukuda, Takanari; Fukuda, Kodai; Tada, Kenichi; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
JAEA-Data/Code 2025-021, 28 Pages, 2026/03
JAEA-Data-Code-2025-021.pdf:1.39MB
Japan Atomic Energy Agency has developed a high-fidelity multi-physics platform JAMPAN for connecting single-physics codes such as a neutronics code and a thermal-hydraulics code. It consists of the HDF5 formatted data container and input/output data handler modules to generate the input file and read the output file of the single-physics codes. Users can easily add or exchange the code by implementing input and output data handler modules for this code. JAMPAN is equipped with interfaces compatible with the neutronics code MVP and the thermal-hydraulics codes JUPITER, ACE-3D, and NASCA, enabling neutronics and thermal-hydraulics coupling calculations to provide reference solutions for core analysis codes. Users can select the thermal-hydraulics code depending on the required calculation accuracy. In addition, the fuel rod properties can be calculated using FEMAXI. This report explains the overview of JAMPAN.
Watanabe, Tomoaki; Aizawa, Naoto*; Chiba, Go*; Tada, Kenichi; Fujita, Tatsuya*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 63(2), p.166 - 186, 2026/02
Times Cited Count：1 Percentile：54.69(Nuclear Science & Technology)A fast burnup calculation method based on neutron spectrum reconstruction is proposed. The method employs a reduced-order model (ROM), constructed using proper orthogonal decomposition (POD) and regression models, to estimate neutron spectra experienced by fuel during burnup. The ROM is built from snapshot data generated through detailed burnup and neutron transport simulations under various conditions. During burnup calculations, the ROM is used to rapidly reconstruct neutron spectra at each burnup step. These reconstructed spectra are then used to compute one-group cross sections from multi-group effective cross sections derived using background cross sections. The proposed method significantly reduces computational time by avoiding repeated neutron transport simulations. Its performance is demonstrated using a PWR UO
fuel pin model. Results show that, with the 6th-order POD, the method predicts nuclide inventories with an average error within 
5% compared to reference Monte Carlo calculations. Error analysis indicates that prediction accuracy is primarily limited by the regression models, rather than by the POD truncation or the multi-group cross section calculations.
Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 63(1), p.3 - 18, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study investigated the impact of nuclear data updates from JENDL-4.0 (J4) to JENDL-5 (J5) on the light-water reactor fuel burnup calculations. Burnup calculations were conducted with J4 and J5 for PWR pin-cell and BWR fuel assembly geometries. The calculation results revealed significant burnup-dependent differences in the neutron multiplication factor (k
). Across the burnup range of 0-50 GWd/t, k
values of J5 were consistently smaller than those of J4 and the difference gradually increased as burnup progressed. Direct sensitivity calculations, in which each nuclide data was replaced from J4 to J5, indicated that updates to the cross-sections of 
U, 
U, and 
Pu and the thermal scattering law data of H in HO notably impacted the k differences. For the BWR assembly geometry containing Gd fuels, large k differences were observed in the burnup range of 10-15 GWd/t. This difference was primarily attributed to updates in the U, 
Gd, and 
Gd cross-sections, and thermal scattering law data of H in HO. Furthermore, we investigated how the nuclear data updates affected the k differences by examining nuclide number densities, the energy-dependent sensitivities, and the neutron spectra.
Tada, Kenichi; Aizawa, Naoto*; Fujita, Tatsuya*; Fukushima, Masahiro; Pyeon, C. H.*
Journal of Nuclear Science and Technology, 63(1), p.1 - 2, 2026/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This document is the preface to "Special Issue on Progressive Reactor Physics for Current and Future Challenges" published in the Journal of Nuclear Science and Technology.
Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki
Mechanical Engineering Journal (Internet), 12(4), p.24-00461_1 - 24-00461_9, 2025/08
JAEA has developed the JAEA Advanced Multi-Physics Analysis platform for Nuclear systems (JAMPAN) to realize high-fidelity neutronics/thermal-hydraulics coupling simulations. We performed a neutronics/thermal-hydraulics coupling simulation for a single BWR fuel assembly in order to confirm that the MVP/JUPITER coupling through JAMPAN is feasible. As a result, we confirmed that the void fraction and the corresponding change in the heat generation distribution are reasonable qualitatively.
Watanabe, Tomoaki; Aizawa, Naoto*; Chiba, Go*; Tada, Kenichi; Yamamoto, Akio*
Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025) (Internet), p.288 - 297, 2025/04
Currently, a major burnup calculation method for the nuclide composition of nuclear fuel conducts neutron transport calculations at each burnup step to account for changes in the neutron spectrum. While this method is highly accurate, the large computational cost of neutron transport calculations can be problematic. Therefore, a fast burnup calculation method based on neutron spectrum reconstruction with the proper orthogonal decomposition (POD) and regression model is investigated. In this method, dimensionality reduction by POD is applied to many neutron fluxes obtained from detailed burnup calculations for various input parameter sets, and regression models are constructed to connect the dimensionality-reduced neutron fluxes and parameters. By substituting arbitrary input parameters to the regression models, the neutron flux is reconstructed and the burnup calculation is performed. This method performs burnup calculations that consider changes in the neutron spectrum based on input conditions without neutron transport calculations. The present method was applied to a PWR UO fuel pin cell model. The results show the nuclide inventory can be calculated with a prediction accuracy within a few percent. In addition, it is found that the calculation error is dominated by the regression models, which implies the further improvement of the regression models leads to improving the accuracy.
Otsuka, Naohiko*; Tada, Kenichi; Cabellos, O.*; Iwamoto, Osamu
Annals of Nuclear Energy, 212, p.110977_1 - 110977_9, 2025/03
Times Cited Count：4 Percentile：57.90(Nuclear Science & Technology)The uranium-233 neutron capture cross section between 3 keV and 1 MeV was evaluated considering the recent new alpha-value measurement performed at the Los Alamos National Laboratory LANCE facility. The obtained capture cross section is systematically lower than the capture cross section in the JENDL-5 library and the reduction is close to 50% around 20 keV. The newly evaluated cross section was validated against 166 criticality experiments chosen from the ICSBEP handbook by performing Monte Carlo neutron transport calculation with the JENDL-5 library, and slight reduction of the chi-square value was achieved by adoption of the newly evaluated capture cross section.
Tada, Kenichi; Kawase, Shoichiro*
Kaku Deta Nyusu (Internet), (140), p.26 - 46, 2025/02
This article summarizes presentations at the IAEA technical meeting on nuclear data retrieval, dissemination, and data portals held in 11-15 November 2024. The purpose of this technical meeting is to discuss nuclear data retrieval, dissemination of data and data portals and to present new developments and future plans. This article explains the overview of presentations in this meeting.
Watanabe, Tomoaki; Suyama, Kenya; Tada, Kenichi; Ferrer, R. M.*; Hykes, J.*; Wemple, C. A.*
Nuclear Science and Engineering, 198(11), p.2230 - 2239, 2024/11
Times Cited Count：2 Percentile：22.05(Nuclear Science & Technology)A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, many essential nuclides for conventional LWR analysis have also been modified based on state-of-the-art evaluations. The new JENDL-5-based CASMO5 library was prepared by replacing as much of the nuclear data of the current CASMO5 ENDF/B-VII.1-based library as possible with JENDL-5. This study verified and validated the new library. Verifications were performed based on the OECD/NEA burnup credit criticality safety benchmark phase III-C, and the calculated k and fuel compositions of the BWR fuel assembly were compared with reported benchmark results. Comparison with the MCNP6.2 result was also performed using the same benchmark model. In addition, the TCA critical experiment and Takahama-3 post-irradiation experiment were used for validation. The results indicate that the new library performs well and is comparable to the ENDF/B-VII.1-based library in predictions of reactivity and fuel compositions for LWR systems.
Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Tada, Kenichi; Kondo, Ryoichi; Nagaya, Yasunobu; Yoshida, Hiroyuki
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/11
We have developed the JAEA Advances Multi-Physics Analysis platform for Nuclear systems (JAMPAN) to realize high-fidelity neutronics/thermal-hydraulics coupling simulations. We will perform MVP/JUPITER coupling simulation for a single BWR fuel assembly in order to confirm that the neutronics/thermal-hydraulics coupling through JAMPAN is feasible. This presentation explains how to send and receive data between MVP and JUPITER through JAMPAN and simulation results.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Tada, Kenichi; Katabuchi, Tatsuya*
Kaku Deta Nyusu (Internet), (139), p.1 - 7, 2024/10
no abstracts in English
Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Konno, Chikara; Kondo, Ryoichi; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*
Journal of Nuclear Science and Technology, 61(6), p.830 - 839, 2024/06
Times Cited Count：14 Percentile：91.83(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.
Tada, Kenichi
Robutsuri No Kenkyu (Internet), (77), 6 Pages, 2024/06
The author participated in the international conference on reactor physics (Physor2024) held in San Francisco, U.S.A. from April 21 to 24, 2024. This article shows the overview of two workshops, i.e., the demonstration of the multi-physics platform Kraken at VTT and MOOSE at INL, the overview of two sections, i.e., "Data Methods, Code Validation" and "Multi-Physics Reactor Simulations and Validation", and the author's impressions of the conference.
Harada, Yoshinari*; Yamaguchi, Hibiki*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi
Transactions of the American Nuclear Society, 130(1), p.758 - 762, 2024/06
The data assimilation was performed using deterministic sampling to selectively reduce uncertainties caused by the thermal neutron scattering in light water. The prompt neutron decay constant 
of the water tank system was used for the data assimilation. The deterministic sampling method was applied to uncertainty quantification and data assimilation for light water thermal neutron scattering law data obtained by the CAB model. The uncertainty quantification results using the deterministic sampling method were comparable to those using the random sampling method.
Tada, Kenichi; Kondo, Ryoichi; Kamiya, Tomohiro; Nagatake, Taku; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
Proceedings of International Conference on Physics of Reactors (PHYSOR 2024) (Internet), p.1488 - 1497, 2024/04
JAEA has developed a new high-fidelity multi-physics platform JAMPAN for connecting single-physics codes such as a neutronics code and a thermal-hydraulics code. It consists of the HDF5 formatted data container and input and output data handler modules to generate the input file and read the output file of the single-physics code. Users can easily add or exchange the code by implementing input and output data handler modules for this code. The first target of JAMPAN is the coupling of neutronics and thermal-hydraulics calculations to provide reference results of core analysis codes. The current version of JAMPAN couples the neutronics code MVP and the thermal-hydraulics codes JUPITER, ACE-3D, and NASCA. Users can select the thermal-hydraulics code depending on the scale of problems to be solved, computational performance, and so on. This presentation explains the overview of JAMPAN and shows the results of the neutronics and thermal-hydraulics coupling calculation.
Tada, Kenichi; Nagaya, Yasunobu; Taninaka, Hiroshi; Yokoyama, Kenji; Okita, Shoichiro; Oizumi, Akito; Fukushima, Masahiro; Nakayama, Shinsuke
Journal of Nuclear Science and Technology, 61(1), p.2 - 22, 2024/01
Times Cited Count：14 Percentile：92.89(Nuclear Science & Technology)The new version of the Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. This paper demonstrates the validation of JENDL-5 for fission reactor applications. Benchmark calculations are performed with the continuous-energy Monte Carlo codes MVP and MCNP and the deterministic code system MARBLE. The benchmark calculation results indicate that the performance of JENDL-5 for fission reactor applications is better than that of the former library JENDL-4.0.
Tada, Kenichi; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 60(11), p.1397 - 1405, 2023/11
Times Cited Count：2 Percentile：13.31(Nuclear Science & Technology)The probability table method is a well-known method for addressing self-shielding effects in the unresolved resonance region. A long computational time is required to generate the probability table. The effective way to reduce the generation time of the probability table is the reduction of the number of ladders. The purpose of this study is the estimation of the optimal number of ladders using the statistical uncertainty in the probability table. To this end, the statistical uncertainty quantification method of the probability table was developed and the convergence behavior of the statistical uncertainty was investigated. The product of the probability table and the average cross section was considered the target of the statistical uncertainty. The convergence rate was affected by the average level spacing and reduced neutron width. The generation time of the probability table was less than half when the input parameter was changed from the number of ladders to the tolerance value.
Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(11), p.1386 - 1396, 2023/11
Times Cited Count：5 Percentile：61.58(Nuclear Science & Technology)The burnup calculations for estimating the nuclide composition of the spent fuel are highly dependent on nuclear data. Many nuclides in the latest version of the Japanese Evaluated Nuclear Data Library JENDL-5 were modified from JENDL-4.0 and the modification affects the burnup calculations. This study confirmed the validity of JENDL-5 in the burnup calculations. The PIE data of Takahama-3 was used for the validation. The effect of modifications of the parameters, e.g., cross sections and fission yields, from JENDL-4.0 to JENDL-5 on the nuclide compositions was quantitatively investigated. The calculation results showed that JENDL-5 has a similar performance to JENDL-4.0. The calculation results also revealed that the modifications of the cross sections of actinide nuclides, fission yields, and thermal scattering low data of hydrogen in HO affected the nuclide compositions of PWR spent fuels.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Tada, Kenichi; Yokoyama, Kenji
Kaku Deta Nyusu (Internet), (136), 6 Pages, 2023/10
no abstracts in English
