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Konno, Chikara
JAEA-Data/Code 2025-019, 70 Pages, 2026/03
JAEA-Data-Code-2025-019.pdf:4.47MB
AMPX format libraries were produced from the evaluated nuclear data library JENDL-5 to make JENDL-5 usable in the US nuclear safety analysis code systems SCALE6.2 and SCALE6.3, which are widely used in Japan. The produced libraries are an AMPX continuous energy library, AMPX multigroup libraries and AMPX covariance libraries. This report explains in detail how to produce the libraries and describes test calculation results for verification of the AMPX libraries.
Kwon, Saerom*; Konno, Chikara; Honda, Shogo*; Kenjo, Shunsuke*; Sato, Satoshi*
Fusion Engineering and Design, 223, p.115548_1 - 115548_8, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In order to evaluate the accuracy of the iron data in the latest nuclear data libraries (FENDL-3.2b, JENDL-5, ENDF/B-VIII.0 and JEFF-3.3) used in the fusion neutron source design, we performed their benchmark tests by using QST/TIARA iron experiment with quasi mono-energy neutrons of 40 and 65 MeV and JAEA/FNS iron experiment with DT neutrons. From the test results, we have found the following issues; (1) The calculation results with FENDL-3.2b underestimate the measured neutron fluxes of the continuous energy range (10-60 MeV) by a factor of 0.6 in the TIARA experiment with 65 MeV neutrons; (2) The calculation results with FENDL-3.2b tend to underestimate the measured neutron flux above 10 MeV by a factor of 0.8 at depth of 70 cm and overestimate the measured ones below 10 keV by a factor of 1.3 up to depth of 40 cm in the FNS experiment. We investigated those issues in detail and clarified their reasons.
Suyama, Kenya
Kaku Deta Nyusu (Internet), (143), p.52 - 62, 2026/02
The Seminar on Nuclear Data holds the longest history as a domestic conference on nuclear data. It has played a vital role as a forum for addressing challenges and identifying new themes in this field, bringing together participants from numerous Japanese universities, research institutions, and industry. With many students attending each session, it offers young researchers who will lead the future an opportunity to present their work and gain exposure to cutting-edge research. As announced at last year's conference, this year's event was held jointly with the ERATO Sekiguchi Three-Nucleon Force Project (TOMOE Project) as the "025 Nuclear Data + TOMOE Project Joint Seminar," aiming to leverage synergies with the latest results in nuclear physics research. While the Seminar on Nuclear Data has been reported, the practical preparation procedures have not been covered extensively in Nuclear Data News. Intended also as handover material, this article covers preparations for the event.
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 H
O 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.
Yanagisawa, Hiroshi; Motome, Yuiko
JAEA-Research 2025-010, 197 Pages, 2025/11
JAEA-Research-2025-010.pdf:3.5MB
For understandings of nuclear criticality risks of TRIGA fuel rods and review of safety measures for handling them, nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of the NSRR fuel rods were re-evaluated with the use of a detailed computational model for the fuel rod. The MVP version 3 code was used with the JENDL libraries including the latest version, JENDL-5, for the re-evaluation. As the criticality characteristics, variations of neutron multiplication factors of the infinite and water-reflected finite systems were examined in detail with parameters of the lattice pitch and density of moderator water. From the results of the re-evaluated criticality characteristics, the minimum critical number of fuel rods for the water-reflected hexagonal shaped lattice system was obtained to be 46.8 
0.2 using the JENDL-5 library. Moreover, the attainability of criticality without the water as moderator and reflector was examined because the zirconium hydride moderator and graphite reflector are equipped with the TRIGA fuel rod. It was found that the criticality is possible to be attained by 115.7 0.6 of the number of fuel rods, which is the smaller number of fuel rods than loaded in the NSRR standard core, even though no water exists.
and -MOX lattice calculationsFujita, Tatsuya
Journal of Nuclear Science and Technology, 62(8), p.731 - 739, 2025/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study estimated the influence of implicit effect on the k-infinity uncertainty in the PWR-UO and -MOX fuel lattice geometries. Firstly, the preliminary investigation was performed, where the influence of implicit effect was roughly estimated based on the sandwich formula using the cross-section (XS) covariance matrix and the sensitivity coefficient. It was confirmed that the influence of implicit effect became large in the fission and (n,
) reactions of heavy nuclides and the change of this dependence was small for the burnup of UO and MOX fuel assemblies. Then, focussing on the heavy nuclides, the influence of implicit effect was compared under several energy group conditions of the XS covariance matrix and neutron transport calculation. For Pu and 
Pu, the noticeable influence of implicit effect was observed in MOX fuel pin-cell geometry. However, increasing the number of energy groups for neutron transport calculations and that of the XS covariance matrix can reduce the influence of implicit effect. Consequently, by appropriately setting the number of energy groups for neutron transport calculations and that of the XS covariance matrix, it became practically possible not to explicitly consider the implicit effect during the random sampling.
Yanagisawa, Hiroshi; Motome, Yuiko
JAEA-Research 2025-001, 99 Pages, 2025/06
JAEA-Research-2025-001.pdf:1.98MB
The detailed computational models for nuclear criticality analyses on the first startup cores of NSRR (Nuclear Safety Research Reactor), which is categorized as a TRIGA-ACPR (Annular Core Pulse Reactor), were created for the purposes of deeper understandings of safety inspection data on the neutron absorber rod worths of reactivity and improvement of determination technique of the reactivity worths. The uncertainties in effective neutron multiplication factor (k
) propagated from errors in the geometry, material, and operation data for the present models were evaluated in detail by using the MVP version 3 code with the latest Japanese nuclear data library, JENDL-5, and the previous versions of JENDL libraries. As a result, the overall uncertainties in k for the present models were evaluated to be in the range of 0.0027 to 0.0029 
k. It is expected that the present models will be utilized as the benchmark on k for TRIGA-ACPR. Moreover, it is confirmed that the overall uncertainties were sufficiently smaller than the values of absorber rod worths determined in NSRR. Thus, it is also considered that the present models are applicable to further analyses on the absorber rod worths in NSRR.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 199(6), p.1029 - 1043, 2025/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 199(1), p.18 - 41, 2025/01
Times Cited Count：2 Percentile：44.79(Nuclear Science & Technology)A series of integral experiments were conducted at FCA of JAEA, simulating LWR cores with a tight lattice cell of highly enriched MOX fuel containing more than 15% fissile plutonium. The three experimental configurations were constructed using foamed polystyrene with different void fractions to clarify the prediction accuracy of neutronic calculation codes and nuclear data among various neutron spectra. The nuclear characteristics measured in the experiments were criticality, moderator void reactivity worths, and sample reactivity worths. The preliminary analyses on experiments were conducted using a deterministic calculation code conventionally used for fast reactors with JENDL-4.0. Most reactivity worth calculations correlated well with the experimental values. Specifically for the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations were performed with MVP3 code by modeling the experimental setup in detail, confirming the validity of the deterministic calculations.
Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu
Nuclear Science and Engineering, 17 Pages, 2025/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Kochiyama, Mami; Murakami, Masashi; Kubota, Shintaro; Sakai, Akihiro
Proceedings of International Conference on Nuclear Decommissioning; Addressing the Past and Ensuring the Future 2023 (Internet), 5 Pages, 2025/00
Japan Atomic Energy Agency has been studied radioactivity evaluation methods with activation calculations as part of those for radioactive waste generated by the decommissioning of research reactors. As part of the study, calculation tools and libraries have been evaluated. Activation calculations were performed using the ORIGEN or ORIGEN-S codes and the cross-section library attached to the SCALE system or JENDL/AD-2017 library, and the results were compared.
Fukahori, Tokio
JAEA-Conf 2024-002, p.6 - 11, 2024/11
The author has been keeping relationship with Japanese Evaluated Nuclear Data Library for around 38 years. During this period, he has been contributing code developments, for example, Preequilibrium Nuclear Reaction Calculation Code (ALICE-F) and Particle and Heavy Ion Transport Code System (PHITS). The author has also been assisting for EXFOR activities and promoting Evaluated Nuclear Data Processing Code (FRENDY) and Multiphase Multicomponent Detailed Thermal Fluid Analysis Code (JUPITER). In this paper, introduced are the outline of the author's works. Also reported are the author's perspective and challenge for the future form of nuclear data.
from 3-GeV protons and 
Hg with the 
Bi(n,xn) reactionsSugihara, Kenta*; Meigo, Shinichiro; Iwamoto, Hiroki; Maekawa, Fujio
JAEA-Conf 2024-002, p.162 - 167, 2024/11
A neutron energy spectrum is important for shielding design at an Accelerator-Driven System facility (1.5-GeV p + Lead Bismuth Eutectic). A similar spectrum can be obtained at J-PARC (3-GeV proton + Hg). To check the validity of the unfolding, the unfolding with the Bi(n,xn) reactions and the response functions (JENDL/HE-2007 and TALYS) was applied. In our poster, we present the derivation of the spectrum and comparison with the spectrum with a Time-of-Flight technique.
Konno, Chikara
JAEA-Conf 2024-002, p.80 - 85, 2024/11
The official ACE files of JENDL-5 were released in December, 2022. The neutron ACE file of JENDL-5 was mainly produced with the FRENDY code, while the data on nuclear heating and damage (heating number, damage production energy) were done with the NJOY2016.65 code modified for JENDL-5. This presentation explains the modified points of NJOY2016.65 and the data on nuclear heating and damage in the neutron ACE file of JENDL-5.
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：1 Percentile：23.55(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.
Okita, Shoichiro; Aoki, Takeshi; Fukaya, Yuji; Tachibana, Yukio
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11
Okita, Shoichiro; Abe, Yutaka*; Tasaki, Seiji*; Fukaya, Yuji
Radioisotopes, 73(3), p.233 - 240, 2024/11
Fujita, Tatsuya
Proceedings of Best Estimate Plus Uncertainty International Conference (BEPU 2024) (Internet), 14 Pages, 2024/05
The uncertainty analysis of PWR depletion test problem on the OECD/NEA/NSC LWR-UAM benchmark Phase II based on JENDL-5 was performed as a preliminary investigation. The random sampling was used to quantify the uncertainty of k-infinity and nuclide inventories, the cross section (XS), the fission product yield (FPY), the decay constant, and the decay branch ratio were randomly perturbed, and several times of SERPENT 2.2.1 calculations were performed. XSs in the ACE file were perturbed by the ACE file perturbation tool using FRENDY with the 56-group covariance matrix generated by NJOY2016.72. The perturbation quantity of independent FPY was evaluated using the FPY covariance matrix prepared in JENDL-5, and the perturbed cumulative FPY was reconstructed based on the relationship between the independent and cumulative FPYs. The decay constant was independently perturbed for each nuclide. To perturb the decay branch ratios, the covariance matrix was generated by applying the generalized least square method and randomly perturbed based on this covariance matrix in the same manner as the independent FPY. In general, the influence due to decay data was an order of magnitude smaller than the influences due to XS and FPY uncertainties. For the uncertainty of k-infinity and transuranic nuclide inventories, the influence due to XS uncertainty was dominant, and that due to FPY and decay data uncertainties was one or a few orders of magnitude smaller. On the other hand, for the uncertainty of FP nuclide inventories, the influence due to FPY uncertainty was almost the same or larger than that due to XS uncertainty. It was also confirmed that the influence due to either XS or FPY uncertainty became different for each FP nuclide. In future studies, the influence due to XS uncertainty on FP nuclides will be discussed because it was not prepared in JENDL-5 and not considered in the present paper.
