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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.
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.
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.
Tada, Kenichi; Ikehara, Tadashi; Ono, Michitaka*; Tojo, Masayuki*
no journal, ,
JAEA develops a nuclear data processing code FRENDY. We can comparison and verification of conventional nuclear data processing code using FRENDY. In this presentation, we focus on the thermal scattering law data. We found some problems of NJOY to process the thermal scattering law data as follows, (1) The generation of input file is complex and we found some inputting error in the official ACE library, (2) The maximum energy of ACE file is not identical to the inputted maximum energy, (3) If user uses iwt=2 option in ACER module, MCNP6.1 cannot treat this generated ACE file appropriately and the calculation will not completed This presentation explains the overview of these problems.
Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Tada, Kenichi
no journal, ,
The multi-group cross section generation function FRENDY/MG is now under development. FRENDY/MG generates a multi-group cross section file from ACE file which is cross section file for a continuous energy Monte Carlo calculation code. Using FRENDY/MG and FRENDY version 1, the multi-group cross section file can be generated.
Ono, Michitaka*; Tojo, Masayuki*; Tada, Kenichi; Iwamoto, Osamu
no journal, ,
The EOLE criticality experiment is analyzed by using the newly released evaluated nuclear data library JENDL-5 for the validation. This presentation reports the impact of the Am-241 modification on the neutronics calculation.
Konno, Chikara
no journal, ,
We examined the ACE files of the Fusion Evaluated Nuclear Data Library FENDL-3.2b in detail and found that damage production energy cross section data and proton ACE files had problems. It was pointed out that the damage production energy cross section data above 20 MeV of FENDL-3.2b from JENDL-4.0/HE and those above a few MeV of FENDL-3.2b from JEFF-3.1.1 were too small, which was solved by replacing them with JENDL-5 and JEFF-3.2 in the next FENDL, respectively. It was also proposed to reproduce the proton ACE files of FENDL-3.2b from JENDL/HE-2007 for the next FENDL with the nuclear data processing code NJOY2016 modified for JENDL-5, because the Monte Carlo code MCNP6.2 could not treat secondary neutron production data in the proton ACE files of FENDL-3.2b from JENDL/HE-2007 adequately.
