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Yee-Rendon, B.; Kondo, Yasuhiro; Tamura, Jun; Meigo, Shinichiro; Maekawa, Fujio; Miyamoto, Ryoichi*; Milas, N.*; Eshraqi, M.*
Proceedings of 22nd International Conference on RF Superconductivity (SRF2025) (Internet), p.58 - 62, 2026/04
The European Spallation Source (ESS) superconducting proton linac is currently undergoing commissioning. During the initial operation phase, the final beam energy will be about 800 MeV, reaching a 2 MW power. High reliability and availability are crucial for the success of the ESS science programs and thus operations will be maintained even with failures of main linac components such as cavities and quadrupoles, as long as ~50 percent of the intended power can be achieved. To this end, we developed beam optics strategies to address failures in the cavities of the superconducting linac. Due to the constraints in the RF cavity amplitudes, we implemented a modified version of standard cavity compensation techniques. The results indicated that this strategy enables beam recovery that meets the beam quality specifications, thereby enhancing the availability of the ESS linac.
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.
Kondo, Ryoichi; Yamamoto, Akio*; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 63(2), p.142 - 153, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The flux distribution tallies using proper orthogonal decomposition, called POD tallies, have been developed to address issues of high-fidelity Monte Carlo simulations. The POD tallies can reduce both dimensionality and statistical error. The present study verifies the applicability of the POD tallies to sub-pin level flux distribution in the two-dimensional C5G7 benchmark. Statistical error estimation is also proposed by applying the circular block bootstrap method to the POD tallies to estimate the statistical error of the flux distribution in a single Monte Carlo calculation. In the verification, the dimensionality of the finely discretized distribution is reduced by more than 90% compared with conventional cell tallies. The statistical error is reduced by more than half as the average value of all tally regions. The proposed approach is confirmed to properly estimate the statistical error of flux distribution considering both the inter-cycle correlation and the correlation between the expansion coefficients of different POD orders.
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.
Kondo, Ryoichi; Yamamoto, Akio*; Endo, Tomohiro*
EPJ Nuclear Sciences & Technologies (Internet), 11, p.21_1 - 21_9, 2025/06
The flux distribution tallies using the proper orthogonal decomposition (POD) called "the POD tallies" have been developed in our previous study. The POD tallies can achieve dimensionality and statistical uncertainty reduction for a finely discretized flux distribution. Some characteristics of the POD tallies, which are left by our previous work, are revealed in the present study. Firstly, the POD tallies with the track length estimator are newly implemented. Since the statistical uncertainty of the POD tallies is reduced compared with the cell tallies, the POD tallies with the track length estimator can obtain the most precise result among the present implantations. Secondly, the basis vectors obtained by the deterministic and the stochastic methods are compared. The statistical uncertainty of the snapshot data invokes the degradation of the extracted basis vectors. This result indicates that the deterministic method might be more efficient for the snapshot calculation. Finally, the impact of the covariances of expansion coefficients on the statistical uncertainty of expanded flux distribution is investigated. The reconstructed statistical uncertainty considering only the variances of the expansion coefficients differs from the reference. This result reveals that the covariances of the expansion coefficients are important to estimate the statistical uncertainty of the local flux in the flux distribution.
Kondo, Ryoichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 61(12), p.1536 - 1545, 2024/12
Times Cited Count：4 Percentile：71.04(Nuclear Science & Technology)In this paper, a new tallying method for a neutron flux distribution using the proper orthogonal decomposition is proposed for dimensionality reduction. The target spatial flux distribution is expanded by orthogonal basis vectors. Expansion coefficients are tallied during the random walk of the Monte Carlo calculation. The orthogonal basis vectors are extracted from the pre-calculated snapshots by the singular value decomposition. The proposed method is verified in the multi-group Monte Carlo calculation with the one-dimensional heterogeneous whole core geometry as a feasibility study. The flux distribution for each of the assemblies and energy groups is expanded by the basis vectors. The fewer basis vectors obtained from snapshots can reconstruct the target distribution well compared with the conventional Legendre polynomials used in the functional expansion tallies. The dimension of the solution in the proposed method is reduced by a factor of twenty compared with the conventional cell tally. In addition, the statistical error is reduced through dimensionality reduction thanks to the methodological feature of the proposed method. The results indicate that the proposed method has the capability of dimensionality reduction to tally the finely discretized flux distribution.
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.
Kondo, Ryoichi; Endo, Tomohiro*; Yamamoto, Akio*
EPJ Web of Conferences, 302, p.04002_1 - 04002_10, 2024/10
Times Cited Count：0 Percentile：0.00(Computer Science, Interdisciplinary Applications)In the recent study, we have developed an efficient flux distribution tallying method in the Monte Carlo calculation toward the high-fidelity, large scale multi-physics simulation. In this method, the proper orthogonal decomposition is applied to the flux distribution tallies. While the tallying method was implemented with the collision estimator in the previous study, the track length estimator is implemented in the present study to obtain the tally with lower statistical error. The implementation of the flux distribution tally with the track length estimator is compared with that of the collision estimator and the normal track length estimator in a one-dimensional problem. The numerical results reveal that the distribution tally using the POD with the track length estimator can obtain a more precise solution compared with that with the collision estimator. Therefore, in terms of the statistical error, the relationship between the distribution tally with track length and collision estimator is similar to that between the conventional track length and collision estimators.
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：13 Percentile：92.36(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; 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.
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：9 Percentile：72.04(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.
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.
Kawase, Keigo; Kando, Masaki; Hayakawa, Takehito; Daito, Izuru; Kondo, Shuji; Homma, Takayuki; Kameshima, Takashi*; Kotaki, Hideyuki; Chen, L. M.*; Fukuda, Yuji; et al.
Nuclear Instruments and Methods in Physics Research A, 637(1, Suppl.), p.S141 - S144, 2011/05
Times Cited Count：7 Percentile：45.61(Instruments & Instrumentation)We report the present status of the sub-MeV X-ray generation via Compton backscattering by using 150-MeV electron beam and the Nd:YAG laser. In particular, we show the result of the X-ray generation experiment and of the laser pulse compression for increasing the X-ray flux.
Kawase, Keigo; Kando, Masaki; Hayakawa, Takehito; Daito, Izuru; Kondo, Shuji; Homma, Takayuki; Kameshima, Takashi*; Kotaki, Hideyuki; Chen, L. M.*; Fukuda, Yuji; et al.
JAEA-Conf 2010-002, p.95 - 98, 2010/06
At the previous symposium in Advanced Photon Research, we proposed and demonstrated the laser pulse compression via stimulated Brillouin scattering (SBS) for increasing the flux of the Compton backscattered X rays. After that, we improved the SBS pulse compression system by introducing the image relay in the laser transport line. As a result, we achieve the stably compressed laser pulse with a duration of 2.1 ns and with an energy of 0.84 J. By installing this system into the Compton backscattered X-ray source, the X-ray flux will be increased 3.2 times for the present system at the KPSI-JAEA.
Kondo, Ryoichi*; Ozawa, Takeshi*; Horie, Junnosuke*; Handa, Hiroyuki*
PNC TJ202 82-03, 76 Pages, 1982/02
A supplemental study on the JOYO's shielding analysis was performed focussing attention on the behavior of neutron and 
-ray fluxes around the detector guide tube of channel a and on the evaluation of the primary sodium activation with use made of the two-dimensional discrete ordinates transport code. It has been found out that secondary rays emitted from a thin guide tube material of SS-304 has a strong effect of enhancing gamma doserates in the guide tube penetrating through a graphite shielding. The effect of the guide tube penetrating a heavy concrete pedestal has been revealed to be a streaming of neutrons and gammas. It can be concluded that the effect of the detector guide tube on the neutron and ray should be taken into account when comparing the calculated values with experimental values. The neutron flux distribution in the full reactor lower axial two dimensional geometry has been calculated and used, together with the results obtained by the upper axial two dimensional a
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.
Kondo, Ryoichi; Yamamoto, Akio*; Endo, Tomohiro*
no journal, ,
We have recently developed the flux distribution tallies using the proper orthogonal decomposition to reduce the dimensionality of the distribution. In the present study, the circular block bootstrap (CBB) method is applied to a statistical uncertainty estimation of the flux distribution taking into account the intergenerational correlation in a Monte Carlo eigenvalue calculation. The effectiveness of the proposed method is verified for a test problem in a one-dimensional slab geometry.
Tada, Kenichi; Kondo, Ryoichi; Kamiya, Tomohiro; Fukuda, Takanari; Ono, Ayako; Nagaya, Yasunobu; Yoshida, Hiroyuki
no journal, ,
JAEA has developed the multi-physics platform JAMPAN. In the previous presentation, we implemented the flow rate calibration function to JAMPAN for the MVP/NASCA coupling calculation of the BWR multi-fuel assembly geometry. Using this function, we performed the whole core coupling calculation using MVP/NASCA. For the whole BWR calculations, we prepared two types of input files. One is the hypothetical whole core geometry consisting of a 9
9 fuel assembly obtained from the OECD/NEA Phase-3C benchmark. The other is the initial loading core of Peach Bottom unit 2 obtained from the OECD/NEA Peach Bottom turbine trip benchmark. We will show the results of both whole-core BWR coupling calculations in the present presentation.
