Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi
Journal of Nuclear Science and Technology, 60(2), p.132 - 139, 2023/02
Times Cited Count：2 Percentile：53.91(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：9 Percentile：84.69(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.
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.
M.M.Sarker*; ; Masukawa, Fumihiro; Naito, Yoshitaka
JAERI-M 93-203, 39 Pages, 1993/10
no abstracts in English
; Okumura, Keisuke; Takano, Hideki; ; *
JAERI 1323, 68 Pages, 1992/01
no abstracts in English
Tada, Kenichi
no journal, ,
The nuclear data processing is very important process to connect the evaluated nuclear data file and nuclear transport calculation code. Because the nuclear data processing is carried out only when new or modified nuclear data file were released, it is difficult to cultivate the specialist of the nuclear data processing. To enhance understanding of the evaluated nuclear data file and nuclear data processing, this lecture explains the overview of them.
Tada, Kenichi
no journal, ,
This presentation explains an overview of the nuclear data processing code FRENDY. The topics of this presentation are an overview of the nuclear data processing, characteristics of FRENDY, new functions implemented in FRENDY version 2, and the future development plan of FRENDY. This presentation also presents an overview of JENDL-5 cross section libraries and future plans of the other cross section library generations.
