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Maruyama, Shuhei
Robutsuri No Kenkyu (Internet), (78), 7 Pages, 2025/03
no abstracts in English
Suyama, Kenya
Kaku Deta Nyusu (Internet), (140), p.13 - 19, 2025/02
A workshop on EXFOR (Exchange Format), a measured nuclear reaction data database, was held in November 2024. This report provides an overview of the workshop and its results.
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.
Shigyo, Nobuhiro*; Furuta, Takuya; Iwamoto, Yosuke
JAEA-Conf 2024-002, 216 Pages, 2024/11
JAEA-Conf-2024-002.pdf:24.29MB
The 2023 Symposium on Nuclear Data was held at Tokai Industry and Information Plaza "iVil" on November 15-17, 2023. The symposium was organized by the Nuclear Data Division of the Atomic Energy Society of Japan (AESJ) in cooperation with Radiation Engineering Division of AESJ, North Kanto Branch of AESJ, Investigation Committee on Nuclear Data in AESJ, Nuclear Science and Engineering Center of Japan Atomic Energy Agency, and High Energy Accelerator Research Organization. In the symposium, tutorials "Overview of the nuclear data processing code, FRENDY version 2" was proposed and held. Two sessions of lectures and discussions were held: "Recent Topics on Nuclear Data and Particle and Heavy Ion Transport code System (PHITS)". In addition, recent research progress on experiments, nuclear theory, evaluation, benchmark, and applications were presented in the poster session. The total number of participants was 108 participants. Each oral and poster presentation was followed by an active question and answer session. This report consists of a total of 36 papers including 17 oral and 19 poster presentations.
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：8 Percentile：89.79(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.
Nakayama, Shinsuke
JAEA-Review 2024-009, 16 Pages, 2024/05
JAEA-Review-2024-009.pdf:1.17MB
Nuclear data is fundamental data for nuclear energy research and development, and its importance has been widely recognized. On the other hand, for future nuclear data research, it is necessary to sort out which types of data (target nuclides, energies, physical quantities, etc.) should be prioritized. Therefore, a "Task Force (TF) for Nuclear Data Roadmap" was established within Investigation Committee on Nuclear Data in the Atomic Energy Society of Japan to discuss a roadmap for future nuclear data research and development. This document reports the results of the discussion in the TF.
Gunji, Satoshi; Araki, Shohei; Izawa, Kazuhiko; Suyama, Kenya
Proceedings of International Conference on Physics of Reactors (PHYSOR 2024) (Internet), p.227 - 236, 2024/04
It is considered that a large amount of fuel debris was generated in the TEPCO's Fukushima Daiichi Nuclear Power Station accident. In particular, the criticality characteristics of the fuel debris, including concrete components, which are products of molten core-concrete interaction (MCCI), have not been well investigated. In this study, to plan physical simulation in critical experiments at the critical assembly using pseudo fuel debris samples including concrete, we evaluated the sensitivity to the effective multiplication factor of the Si and Ca cross sections in the concrete-simulant sample based on the results of elemental analysis of the prototype. These sensitivity calculations were carried out for each sample loading method and composition. We focused on the energy profile of the sensitivity of the 
Ca capture reaction and confirmed that the shape of the sensitivity energy profile changed depending on the sample compositions and neutron moderation conditions. We could know the characteristics of each experimental method by clarifying the trends of sensitivity obtained in different experimental cases. It was found that increasing the amount of concrete in the samples and changing the neutron moderation conditions in the experimental core configurations produced similar changes in the shape of the sensitivity energy profile. This result shows the possibility of reproducing the characteristics of MCCI products through practical critical experiments using concrete materials that do not contain fissile materials.
Iwamoto, Hiroki
JAEA-Conf 2023-001, p.40 - 45, 2024/02
no abstracts in English
Ra nuclear data library and its impact on the production amount of 
Ac via the Ra (n,2n) reactionSasaki, Yuto; Iwamoto, Nobuyuki; Takaki, Naoyuki*; Maeda, Shigetaka
Journal of Nuclear Science and Technology, 61(2), p.251 - 260, 2024/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Ac is a promising alpha emitter for targeted alpha therapy. However, the current capability of Ac supply is limited to approximately 63 GBq/y, primarily relying on the natural source of 
Th stocked at a few institutes. Therefore, alternative Ac production methods are highly desired. The research and development of Ac and its parent nuclide production methods using accelerators and reactors to target Ra and 
Th are actively pursued worldwide. Hence, the authors focused on the Ra(n,2n) Ra method using fast neutron spectra as an application of the experimental fast reactor Joyo. This study investigated the status of nuclear data libraries for Ra, an essential target for Ac production, and evaluated the impact of different nuclear data libraries on the amount of Ac produced. Consequently, cross-sections with covariance data were stored in TENDL-2021, JEFF-3.3, and EAF-2010 but not in ENDF or JENDL, the major nuclear data libraries. Furthermore, no consistency occurred among the respective nuclear data, and the Ra production amount varied.
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：12 Percentile：96.41(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.
Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 61(1), p.31 - 43, 2024/01
Times Cited Count：3 Percentile：46.61(Nuclear Science & Technology)This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g., reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.
Pb and 
Bi in proton-induced reactions near 100 MeVIwamoto, Hiroki; Meigo, Shinichiro; Satoh, Daiki; Iwamoto, Yosuke; Ishi, Yoshihiro*; Uesugi, Tomonori*; Yashima, Hiroshi*; Nishio, Katsuhisa; Sugihara, Kenta*; 
elik, Y.*; et al.
Nuclear Instruments and Methods in Physics Research B, 544, p.165107_1 - 165107_15, 2023/11
Times Cited Count：4 Percentile：73.39(Instruments & Instrumentation)The lack of double-differential cross-section (DDX) data for neutron production below the incident proton energy of 200 MeV hinders the validation of spallation models in technical applications, such as research and development of accelerator-driven systems (ADSs). The present study aims to obtain experimental DDX data for ADS spallation target materials in this energy region and identify issues related to the spallation models by comparing them with the analytical predictions. The DDXs for the (
) reactions of Pb and Bi in the 100-MeV region were measured over an angular range of 30
to 150 using the time-of-flight method. The measurements were conducted at Kyoto University utilizing the FFAG accelerator. The DDXs obtained were compared with calculation results from Monte Carlo-based spallation models and the evaluated nuclear data library, JENDL-5. Comparison between the measured DDX and analytical values based on the spallation models and evaluated nuclear data library indicated that, in general, the CEM03.03 model demonstrated the closest match to the experimental values. Additionally, the comparison highlighted several issues that need to be addressed in order to improve the reproducibility of the proton-induced neutron-production DDX in the 100 MeV region by these spallation models and evaluated nuclear data library.
Gunji, Satoshi; Araki, Shohei; Watanabe, Tomoaki; Fernex, F.*; Leclaire, N.*; Bardelay, A.*; Suyama, Kenya
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 9 Pages, 2023/10
Institut de radioprotection et de s
ret
nuclaire (IRSN) and Japan Atomic Energy Agency (JAEA) have a long-standing partnership in the field of criticality safety. In this collaboration, IRSN and JAEA are planning a joint experiment using the new STACY critical assembly, modified by JAEA. In order to compare the codes (MVP3, MORET6, etc.) and nuclear data (JENDL and JEFF) used by both institutes in the planning of the STACY experiment, benchmark calculations of the Apparatus B and TCA, which are critical assemblies once owned by both institutes, benchmarks from the ICSBEP handbook and the computational model of the new STACY were performed. Including the new STACY calculation model, the calculations include several different neutron moderation conditions and critical water heights. There were slight systematic differences in the calculation results, which may have originated from the processing and/or format of the nuclear data libraries. However, it was found that the calculated results, including the new codes and the new nuclear data, are in good agreement with the experimental values. Therefore, there are no issues to use them for the design of experiments for the new STACY. Furthermore, the impact of the new TSL data included in JENDL-5 on the effective multiplication factor was investigated. Experimental validation for them will be completed by critical experiments of the new STACY by both institutes.
Tada, Kenichi
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 8 Pages, 2023/10
The number of energy grids of the thermal neutron scattering law data has a large impact on the data size of a cross section file of continuous energy Monte Carlo calculation codes. The optimization of the number of energy grids is an effective way to reduce the data size. This study developed the linearization method of the thermal neutron scattering cross section to optimize the number of energy grids and the linearization function was implemented in the nuclear data processing code FRENDY. The linearization process which is used in the resonance reconstruction and the Doppler broadening was adopted. The criticality benchmarks which use ZrH as the moderator were calculated to estimate the impact of the difference of the energy grids on neutronics calculations. The calculation results indicate that the linearization of the thermal neutron scattering cross section improves the prediction accuracy of neutronics calculations.
Tada, Kenichi
Kaku Deta Nyusu (Internet), (135), p.1 - 10, 2023/06
This article summarizes presentations at the IAEA technical meeting on nuclear data processing. In this technical meeting, the current development status of nuclear data processing codes and comparisons of the processing results using these codes were presented.
Tada, Kenichi; Endo, Tomohiro*
EPJ Web of Conferences, 284, p.14013_1 - 14013_4, 2023/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The self-shielding effect in the unresolved resonance region has a large impact on the fast- and intermediate-spectrum reactors. The probability table method is widely used for continuous-energy Monte Carlo calculation codes to treat the effect. In this method, a table provides the probability distribution of the cross-section for a nuclide in the given energy grid points. The table is generated by averaging with a lot of "ladders" which represent pseudo resonance structures. Though many nuclear data processing codes require the number of ladders as an input parameter to generate the probability table, an optimal number of ladders has not been investigated. Our previous study revealed that the suitable number of ladders depends on the nuclide and its resonance parameters. This result indicates that it is very difficult for users to find the optimal number of ladders. We developed the calculation method of the statistical uncertainty for the probability table generation.
Iwamoto, Hiroki; Nakano, Keita; Meigo, Shinichiro; Satoh, Daiki; Iwamoto, Yosuke; Sugihara, Kenta*; Nishio, Katsuhisa; Ishi, Yoshihiro*; Uesugi, Tomonori*; Kuriyama, Yasutoshi*; et al.
EPJ Web of Conferences, 284, p.01023_1 - 01023_4, 2023/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)For accurate prediction of neutronic characteristics for accelerator-driven systems (ADS) and a source term of spallation neutrons for reactor physics experiments for the ADS at Kyoto University Critical Assembly (KUCA), we have launched an experimental program to measure nuclear data on ADS using the Fixed Field Alternating Gradient (FFAG) accelerator at Kyoto University. As part of this program, the proton-induced double-differential thick-target neutron-yields (TTNYs) and cross-sections (DDXs) for iron, lead, and bismuth have been measured with the time-of-flight (TOF) method. For each measurement, the target was installed in a vacuum chamber on the beamline and bombarded with 107-MeV proton beams accelerated from the FFAG accelerator. Neutrons produced from the targets were detected with stacked, small-sized neutron detectors for several angles from the incident beam direction. The TOF spectra were obtained from the detected signals and the FFAG kicker magnet's logic signals, where gamma-ray events were eliminated by pulse shape discrimination. Finally, the TTNYs and DDXs were obtained from the TOF spectra by relativistic kinematics. The measured TTNYs and DDXs were compared with calculations by the Monte Carlo transport code PHITS with its default physics model of INCL version 4.6 combined with GEM and those with the JENDL-4.0/HE nuclear data library.
Tada, Kenichi
Robutsuri No Kenkyu (Internet), (75), 13 Pages, 2023/03
In addition to nuclear data processing, FRENDY has various functions such as editing nuclear data and plotting cross section data. This document introduces these functions.
Tada, Kenichi
Shahei Kaiseki No V&V Gaidorain Sakutei Ni Mukete, p.11 - 16, 2023/03
An overview of the nuclear data processing code FRENDY is introduced for shielding calculation code users who are not familiar with FRENDY. This paper explains the nuclear data processing flow in FRENDY, the purpose of use, input examples, verification and validation reports, and so on.
