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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：88.57(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.
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.
Konno, Chikara
JAEA-Conf 2023-001, p.143 - 146, 2024/02
I modified NJOY2016.67 to produce photonuclear ACE files which can be used in MCNP6.2 and PHITS3.27 and produced the ACE file of the JENDL-5 photonuclear sub-library. Simple test calculations with the produced ACE file supported that the produced ACE file had no serious problems.
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.
Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi
Journal of Nuclear Science and Technology, 60(2), p.132 - 139, 2023/02
Times Cited Count：3 Percentile：41.50(Nuclear Science & Technology)A new multi-group neutronics analysis sequence ACE-FRENDY-CBZ is proposed. This sequence is free from uses of any application libraries; with the ACE files as the starting point, multi-group cross section data of media comprising a target system are calculated with the FRENDY code, and multi-group neutron transport calculations are performed with modules of the CBZ code system. The ACE-FRENDY-CBZ sequence was tested against the eight fast neutron systems, and good agreement with the reference Monte Carlo results was obtained within 30 pcm differences in the bare systems and the thorium-reflected system, and approximately 100 pcm differences in the uranium-reflected systems. The use of the current-weighted total cross sections in the multi-group neutron transport calculations had non-negligible impacts over 100 pcm on k-eff, and the calculations with the current-weighted total cross sections systematically underestimated k-eff in the uranium-reflected systems.
Yamamoto, Akio*; Tada, Kenichi; Chiba, Go*; Endo, Tomohiro*
Journal of Nuclear Science and Technology, 58(11), p.1165 - 1183, 2021/11
Times Cited Count：15 Percentile：84.25(Nuclear Science & Technology)The multi-group cross section generation capability for neutrons is implemented in the FRENDY nuclear data processing code. ACE-formatted files are used as the source of nuclear data instead of ENDF-formatted files since FRENDY already has the capability to generate pointwise cross sections in the ACE format. Verification calculations of the newly implemented capability are carried out through the comparison with the NJOY nuclear data processing code. Cross section generations for all nuclides in JENDL-4.0, -4.0u, -5
4, ENDF/B-VII.1, -VIII.0, JEFF-3.3, and TENDL-2019 are carried out without unexpected processing issue, except for Pu-238 of TENDL-2019 that includes inconsistent data. The verification results indicate that the multi-group cross sections generated by FRENDY are consistent with those generated by NJOY or the calculation results by MCNP.
Iwamoto, Osamu
JAEA-Conf 2020-001, p.11 - 16, 2020/12
Fleming, M.*; Chadwick, M.*; Brown, D.*; Capote, R.*; Ge, Z.*; Herman, M.*; Ignatyuk, A.*; Ivanova, T.*; Iwamoto, Osamu; Koning, A.*; et al.
EPJ Web of Conferences, 239, p.15003_1 - 15003_5, 2020/09
Times Cited Count：5 Percentile：93.87(Nuclear Science & Technology)Iwamoto, Osamu; Iwamoto, Nobuyuki; Shibata, Keiichi; Ichihara, Akira; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke
EPJ Web of Conferences, 239, p.09002_1 - 09002_6, 2020/09
Times Cited Count：60 Percentile：99.94(Nuclear Science & Technology)Tada, Kenichi; Kunieda, Satoshi; Nagaya, Yasunobu
Proceedings of 2019 IEEE Nuclear Science Symposium and Medical Imaging Conference (NSS/MIC 2019), Vol.1, p.48 - 49, 2020/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Nuclear data processing is an important interface between evaluated nuclear data library and neutronics transport codes. JAEA developed a new nuclear data processing code FRENDY in order to process the evaluated nuclear data library JEND. JAEA released FRENDY version 1 as an open source software under the 2-clause BSD license. The current version of FRENDY can generate ACE formatted files for continuous energy Monte Carlo calculation codes such as MCNP and PHITS. It uses the same processing method as NJOY because the implementation of the conventional method is an important step to develop the new code. For verification, we compared the processing results of FRENDY with those of NJOY and confirmed FRENDY provided us with almost the same outputs as NJOY. In this presentation, we explains the overview and a future plan of FRENDY.
Tada, Kenichi; Iwamoto, Osamu
Proceedings of 2019 IEEE Nuclear Science Symposium and Medical Imaging Conference (NSS/MIC 2019), Vol.2, p.1622 - 1624, 2020/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)JAEA has published the evaluated nuclear data library JENDL to improve the prediction accuracy of nuclear calculations. JENDL is now one of the most famous evaluated nuclear data libraries in the world. This presentation explains the recent activity of the JENDL project and overview of the next version of general-purpose file JENDL-5. Nuclear calculation codes cannot treat the evaluated nuclear data library. This presentation also explains the nuclear data processing system FRENDY which is used to generate cross section library for a nuclear calculation code.
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.
Iwamoto, Osamu
JAEA-Conf 2018-001, p.87 - 91, 2018/12
Status and plan of JENDL will be presented. After the release of JENDL-4.0 in 2010, six special purpose files have been developed. Four of them were already released and two are under preparation for the release. New decay and yield data for fission products were released as JENDL/FPD-2011 and JENDL/FPY-2011 in 2011, respectively. JENDL-4.0/HE released in 2015 includes proton and neutron induced reaction data up to 200 MeV. Comprehensive decay data were released as JENDL/DDF-2015 which contains data for 3,237 nuclides. New photonuclear reaction data JENDL/PD-2016 and an activation file JENDL/AD-2017 are under preparation for release. Regarding general purpose file, two activities are in progress. One is JENL-4.0u which is created for maintenance of JENDL-4.0 and the other is development of next version of JENDL. For the next JENDL, evaluation for light nuclei and structure material are in progress. It is planed that next version of JENDL will be JENDL-5 which contains nuclear data for all nuclei having natural abundance. Addition of covariance data will be one of the main targets.
Hashimoto, Shintaro; Sato, Tatsuhiko; Iwamoto, Yosuke; Ogawa, Tatsuhiko; Furuta, Takuya; Abe, Shinichiro; Niita, Koji*
Kaku Deta Nyusu (Internet), (120), p.26 - 34, 2018/06
Particle and heavy-ion transport code system PHITS has been used for calculations of radiation shielding in accelerator facilities. PHITS describes physical phenomena induced by radiation as combination of transport and collision processes. The collision process including nuclear reactions is simulated by the three-step calculation: a generation of a reaction, pre-equilibrium, and compound processes. In the simulation, many physics models are used. This report explains roles of the models in PHITS and shows their developments we recently performed.
Tada, Kenichi; Kosako, Kazuaki*; Yokoyama, Kenji; Konno, Chikara
Nihon Genshiryoku Gakkai-Shi ATOMO
, 60(3), p.168 - 172, 2018/03
The neutronics calculation codes cannot treat the evaluated nuclear data file directly. The nuclear data processing is required to use the nuclear data file in the neutronics calculation codes. The nuclear data processing is not just a converter but also many processes to evaluate the physical values for the neutronics calculation codes. In this paper, we describe the overview of the nuclear data processing and validation of the nuclear data.
Suyama, Kenya; Yokoyama, Kenji
Kaku Deta Nyusu (Internet), (119), p.38 - 47, 2018/02
We have developed numerous neutronics calculation codes in Japan. However, development of the one-point burnup calculation code which replaces the still widely used ORIGEN2 code has not been successful. The one point burnup code is indispensable to evaluate the characteristics of the used nuclear fuel increasing in Japan, and it uses all evaluated nuclear data including the fission yield and decay data as well as cross section data. It means that it could be the Killer Application in the field of the nuclear data and neutronics code. This report describes the necessity of the one point burnup calculation code development in Japan and required function and performance which have been considered by authors.
Iwamoto, Osamu; Shibata, Keiichi; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Ichihara, Akira; Nakayama, Shinsuke
EPJ Web of Conferences, 146, p.02005_1 - 02005_6, 2017/09
Times Cited Count：1 Percentile：58.02(Nuclear Science & Technology)Tamai, Hiroshi; Okubo, Ayako; Kimura, Yoshiki; Shinohara, Nobuo; Tazaki, Makiko; Shimizu, Ryo; Suda, Kazunori; Tomikawa, Hirofumi
Proceedings of INMM 58th Annual Meeting (Internet), 6 Pages, 2017/07
Nuclear forensics is a technical measure to analyse and collate samples of illegally used nuclear materials, etc., to clarify their origins, routes, etc. and contribute to criminal identifications. Close collaboration with police and judicial organizations is essential. The national response framework is being built up with international cooperation. Discussions on promoting technical capability and regional cooperation are presented.
