Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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, Yasuhiro; Chimura, Motoki; Cicek, E.*; Fukui, Yuji*; Futatsukawa, Kenta*; Fuwa, Yasuhiro; Goto, Rikuto*; Hirano, Koichiro; Ishikawa, Masaki*; Ito, Takashi; et al.
Proceedings of 22nd Annual Meeting of Particle Accelerator Society of Japan (Internet), p.124 - 129, 2026/03
Various upgrade plans of the J-PARC linac are underway, and repetition-rate doubling project is one of the most important part. Currently, the linac supplies a 400~MeV H
ion beam to the 3~GeV synchrotron (RCS) at a repetition rate of 25~Hz. However, to deliver the beam to both RCS and proton-beam irradiation facility under planning, it is necessary to raise the repetition rate from 25~Hz to 50~Hz. Prior to construction of the beamline to the irradiation facility, we conducted a verification of the 50~Hz operation of all the linac apparatus simultaneously within the current operational limits in May 2025. And also, the verification of the 50~Hz operation of the particle counter system, which is one of the most important safety device of the J-PARC accelerator, was conducted. As a result, any significant problems were not found out, and it is now ready for the upcoming 50~Hz-beam acceleration demonstration. This paper describes this details of the demonstration of the 50~Hz operation, without beam acceleration, of the J-PARC linac.
In
( = Co, Rh, Ir)Kurihara, Ryosuke*; Hirose, Yusuke; Matsui, Kazuki*; Miyake, Atsushi*; Tsunoda, Ryoma*; Kondo, Masaki*; Settai, Rikio*; Akatsu, Mitsuhiro*; Nemoto, Yuichi*; Yaguchi, Hiroshi*; et al.
JJAP Conference Proceedings (Internet), 12, p.011006_1 - 011006_6, 2026/02
no abstracts in English
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.
ion beam at J-PARC frontendShibata, Takanori*; Nammo, Kesao*; Shinto, Katsuhiro; Okoshi, Kiyonori; Kawai, Isao*; Kitamura, Ryo; Morishita, Takatoshi; Kondo, Yasuhiro; Sato, Yoichi*
Proceedings of 16th International Particle Accelerator Conference (IPAC25) (Internet), p.514 - 516, 2025/11
no abstracts in English
Liu, Y.*; Otani, Masashi*; Miyao, Tomoaki*; Nakazawa, Yuga*; Shibata, Takanori*; Nammo, Kesao*; Fang, Z.*; Futatsukawa, Kenta*; Fukui, Yuji*; Mizobata, Satoshi*; et al.
Proceedings of 16th International Particle Accelerator Conference (IPAC25) (Internet), p.855 - 857, 2025/11
Sato, Rika; Kondo, Toshiki; Umeda, Ryota; Kikuchi, Shin; Yamano, Hidemasa
Progress in Nuclear Science and Technology (Internet), 8, p.137 - 142, 2025/09
In a sodium-cooled fast reactor (SFR) coupled to thermal energy storage (TES) system, the reaction between nitrate molten salt as thermal energy storage medium and sodium (Na) as reactor coolant might occur under postulated accidental conditions. Thus, the reaction behavior of Na-nitrate molten salt is one of the important phenomena in terms of safety assessment of the SFR with TES system. In this study, reaction experiments on Na-solar salt were performed. It was found that Na-solar salt reaction occurred after the NaNO
-KNO eutectic melting. Based on the measured reaction temperature, the kinetic parameters and rate constant were obtained and compared with the sodium-water reaction. From the results of kinetic analysis, it could be assumed that Na-solar salt reaction occurs in the time frame of the accident such as the failure of heat transfer tube of sodium-molten salt heat exchanger.
Iketani, Shotaro; Suzuki, Takeshi; Yokobori, Tomohiko; Sugawara, Satoshi; Yokota, Akira; Kikuchi, Genta; Muraguchi, Yoshinori; Kitahara, Masaru; Seya, Manato; Kurosawa, Tsuyoshi; et al.
JAEA-Technology 2025-001, 169 Pages, 2025/08
JAEA-Technology-2025-001.pdf:14.22MB
The radioactive waste treatment facilities at the Nuclear Science Research Institute includes the Radioactive Waste Treatment Facility No. 3, Waste Size Reduction and Storage Facility, and Waste Volume Reduction Facility. These three facilities come under the purview of the Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors, and are included under Class C of the act based on the seismic requirements specified in the Act. We assessed the seismic capacity of these three radioactive waste treatment facilities based on the current Building Standards Act, to verify whether they comply with the new regulatory requirements enforced by the Nuclear Regulation Authority (NRA) in the aftermath of the 2011 nuclear accident at the Fukushima Daiichi Nuclear Power Station operated by the Tokyo Electric Power Company. We found that the allowable stress of a few structural members used in the construction of the facilities did not meet the regulatory requirements. After studying the approval granted by the NRA for the construction plans, including the design and construction methods (design and construction plans) of the three facilities on March 5, 2021, we made aseismic reinforcement at these facilities between 2021 and 2022. This report presents an overview of the seismic design of these facilities and an outline of the aseismic reinforcement conducted, management system existing, safety measures adopted, and the preoperational inspections conducted at these facilities.
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：66.28(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.
Yamano, Hidemasa; Takano, Kazuya; Kurisaka, Kenichi; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Sato, Rika; Shirakura, Shota*
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. This paper describes the effect of sodium-molten salt heat transfer tube failure in addition to the project overview and progress.
Kikuchi, Shin; Sato, Rika; Kondo, Toshiki; Umeda, Ryota; Yamano, Hidemasa
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06
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：15 Percentile：92.34(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.
Ishida, Tsuyoshi; Nakashima, Shinichi; Kondo, Shinji; Hayashibara, Kenichi; Yamada, Shigeki*; Okamoto, Ryo*; Nakamura, Hironobu
Dai-44-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2023/11
no abstracts in English
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
