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Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 61(1), p.31 - 43, 2024/01
Times Cited Count：0 Percentile：0.01(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：0 Percentile：0.02(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.
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.
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
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：0.21(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：0.21(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; Nagaya, Yasunobu; Taninaka, Hiroshi; Yokoyama, Kenji; Okita, Shoichiro; Oizumi, Akito; Fukushima, Masahiro; Nakayama, Shinsuke
Journal of Nuclear Science and Technology, 21 Pages, 2023/04
Times Cited Count：6 Percentile：98.92(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
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.
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.
Nakayama, Shinsuke
Kaku Deta Nyusu (Internet), (133), p.88 - 99, 2022/10
The content of the paper that received the Paper Award of Atomic Energy Society of Japan in 2021 is outlined. Although the use of deuteron accelerator-based neutron sources has been proposed in various fields, deuteron nuclear database accurate enough to be applied to the design study of such neutron sources had not been developed. Under these situations, we had developed a deuteron nuclear database, JENDL/DEU-2020. It contains evaluated deuteron nuclear data for light nuclei (
Li, 
Be, 
C), which are candidates for deuteron beam irradiation targets of the neutron sources. Evaluation of JENDL/DEU-2020 was performed by using the code system DEURACS with further modifications. In order to validate the accuracy of the database, simulations using the particle transport code were performed under various conditions with different target nuclides and incident deuteron energies, and the results were compared with the available experimental data. As a result, it was found that JENDL/DEU-2020 significantly improves the prediction accuracy of experimental data under a wider range of conditions than other nuclear reaction databases or the nuclear reaction models implemented in transport calculation codes.
Okumura, Keisuke
Kaku Deta Nyusu (Internet), (133), p.63 - 67, 2022/10
no abstracts in English
Yokoyama, Kenji; Taninaka, Hiroshi
Kaku Deta Nyusu (Internet), (132), p.25 - 33, 2022/06
This article explains the results of integral test of JENDL-5 by benchmark analysis in fast reactor system, which were presented in a special session of the 2022 Spring Annual Meeting of the Atomic Energy Society of Japan (AESJ). The latest version of Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. In order to confirm the applicability of JENDL-5 to the fast reactor system, we conducted a set of benchmark analysis using the integral experiment data included in the fast reactor nuclear design database which is being developed by JAEA. With respect to major nuclear characteristics in the standard fast reactor system, it was confirmed that the ratios of analysis result and experimental result (C/E values) based on JENDL-5 were almost the same as those of JENDL-4.0. In the special session, the results of sensitivity analysis were reported. Since the results have been described in the proceedings of the AESJ meeting, we add the results of the versions under development of JENDL-5 and discuss their relationship with the reported results of sensitivity analysis.
Tada, Kenichi; Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*
Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05
Nuclear data processing is an important interface between an evaluated nuclear data library and nuclear transport calculation codes. JAEA has developed a new nuclear data processing code FRENDY from 2013. FRENDY version 1 generates ACE files which are used for the continuous-energy Monte Carlo codes including PHITS, Serpent, and MCNP; it was released as an open-source software under the 2-clause BSD license in 2019. After FRENDY version 1 was released, many functions are developed: the multi-group neutron cross-section library generation, the statistical uncertainty quantification for the probability tables for unresolved resonance cross-section, the perturbation of the ACE file, and the modification of the ENDF-6 formatted nuclear data file, etc. We released FRENDY version 2 including these functions. This presentation explains the overview of FRENDY and features of the new functions implemented in FRENDY version 2.
Ono, Michitaka*; Tojo, Masayuki*; Tada, Kenichi; Yamamoto, Akio*
Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 9 Pages, 2022/05
In this paper, nuclear calculations were performed using the ACE files and the multigroup libraries created by both FRENDY and NJOY, and the impacts on the neutronics characteristics due to nuclear data processing were investigated using those libraries. MCNP was used to compare the ACE files by calculating many benchmark problems including ICSBEP and it was confirmed that the k-eff values are generally agreed with each other within the range of statistical errors. The multigroup cross sections are verified by the BWR design codes LANCR/AETNA through calculation of a commercial-grade BWR5 equilibrium core loaded with 9
9 fuels. These results indicate that fuel assembly and core characteristics are consistent with each other. From the above investigations, it was confirmed that FRENDY can provide comparable continuous/multi-group neutron cross sections with NJOY.
Imai, Nobuaki*; Otsu, Hideaki*
JAEA-Conf 2021-001, 236 Pages, 2022/03
JAEA-Conf-2021-001.pdf:48.06MB
The 2020 Symposium on Nuclear Data was held on-site at RIBF Conference Hall in RIKEN Wako campus on November 26 to 27, 2020, combined with on-line connection conference. The symposium was organized by the Nuclear Data Division of the Atomic Energy Society of Japan (AESJ) in cooperation with Sigma Investigative Advisory Committee of AESJ, Nuclear Science and Engineering Center of Japan Atomic Energy Agency (JAEA), RIKEN Nishina Center, Center for Nuclear Study, The University of Tokyo (CNS), KEK Wako Nuclear Science Center (WNSC), School of Science, The University of Tokyo. In the symposium, six sessions were proposed and held: "Reactor theory and experiments" as a tutorial, "Accelerator Facilities and Nuclear Data", "Deep/Machine Learning and Nuclear Physics, Application to Nuclear Data", "Nuclear Medicine and Nuclear Pharmacy", "Fission, Heavy Ion Nuclear Spectroscopy", and "Nuclear Reaction Data" as lecture and discussion sessions. 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 119, of which 62 were on-site participants. Each oral and poster presentation was followed by an active question and answer session. This report consists of total 40 papers including 15 oral and 25 poster presentations.
Iwamoto, Hiroki
JAEA-Conf 2021-001, p.83 - 87, 2022/03
We have developed a method to generate nuclear data using Gaussian process regression [1], which is one of the machine learning technique. This method generates nuclear data by treating measured data as the training data in machine learning. Since Gaussian process regression is based on nonparametric Bayesian inference, the generated nuclear data are expressed as a predictive distribution including uncertainty information. In this presentation, the basics of the Gaussian process model, some examples of the application to nuclear data generation, and other related topics will be presented. [1] H. Iwamoto, "Generation of nuclear data using Gaussian process regression", Journal of Nuclear Science and Technology, 50:8, 932-938 (2020).
Iwamoto, Hiroki; Iwamoto, Osamu; Kunieda, Satoshi
Journal of Nuclear Science and Technology, 59(3), p.334 - 344, 2022/03
Times Cited Count：4 Percentile：56.94(Nuclear Science & Technology)A hybrid nuclear data estimator (G-HyND) based on a machine learning technique with Gaussian processes (GP) was developed. G-HyND estimates cross-sections from a hybrid training dataset composed of an experimental dataset and an analytical dataset based on a nuclear physics model, and generates the cross-section datasets including the dataset's uncertainty information. It was demonstrated that an experimental dataset and a physics model-based analytical dataset perform a complementary role in nuclear data generation, and that the generated nuclear data from the hybrid training dataset are more reasonable than only those from the experimental dataset. Furthermore, solutions for two inherent GP problems, i.e., overfitting and computational cost, are presented within the G-HyND framework.
