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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.

Tada, Kenichi; Kondo, Ryoichi; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 60(6), p.624 - 631, 2023/06

Times Cited Count：1 Percentile：40.11(Nuclear Science & Technology)The sensitivity analysis and the uncertainty quantification have an important role in improving the evaluated nuclear data library. The current computational performance enables us to the sensitivity analysis and uncertainty quantification using the continuous energy Monte Carlo calculation code. The ACE file perturbation tool was developed for these calculations using modules of FRENDY. This tool perturbs the microscopic cross section, the number of neutrons per fission, and the fission spectrum. The uncertainty quantification using the random sampling method is also available if the user prepares the covariance matrix. The uncertainty of the k-effective using the perturbation tool was compared to the current sensitivity analysis codes SCALE/TSUNAMI and MCNP/KSEN. The comparison results indicated that the random sampling method using this tool accurately estimates the uncertainty of k-effective.

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

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.

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; Nagaya, Yasunobu

JAEA-Data/Code 2022-009, 208 Pages, 2023/02

The nuclear data processing code has an important role to connect evaluated nuclear data libraries and neutronics calculation codes. Japan Atomic Energy Agency (JAEA) has developed the nuclear data processing code FRENDY since 2013 to generate cross section files from evaluated nuclear data libraries, such as JENDL, ENDF/B, JEFF, and TENDL. The first version of FRENDY was released in 2019. FRENDY version 1 generates ACE files which are used for continuous energy Monte Carlo codes such as PHITS, Serpent, and MCNP. FRENDY version 2 generates multi-group neutron cross-section files from ACE files. The other major improvements are as follows: (1) uncertainty quantification for the probability tables of the unresolved resonance cross-section; (2) perturbation of the ACE file for the uncertainty quantification using a continuous Monte Carlo code; (3) modification of the ENDF-6 formatted nuclear data file. This report describes an overview of the nuclear data processing methods and input instructions for FRENDY.

Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi

Journal of Nuclear Science and Technology, 60(2), p.132 - 139, 2023/02

Times Cited Count：0 Percentile：0.01(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.

Tada, Kenichi; Endo, Tomohiro*

Journal of Nuclear Science and Technology, 9 Pages, 2023/00

Times Cited Count：0 Percentile：0.02(Nuclear Science & Technology)The probability table method is a well-known method for addressing self-shielding effects in the unresolved resonance region. A long computational time is required to generate the probability table. The effective way to reduce the generation time of the probability table is the reduction of the number of ladders. The purpose of this study is the estimation of the optimal number of ladders using the statistical uncertainty in the probability table. To this end, the statistical uncertainty quantification method of the probability table was developed and the convergence behavior of the statistical uncertainty was investigated. The product of the probability table and the average cross section was considered the target of the statistical uncertainty. The convergence rate was affected by the average level spacing and reduced neutron width. The generation time of the probability table was less than half when the input parameter was changed from the number of ladders to the tolerance value.

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

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.

Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Tada, Kenichi

Nuclear Science and Engineering, 196(11), p.1267 - 1279, 2022/11

Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The resonance upscattering effect (the thermal agitation effect) is incorporated in the generation capability of multi-group neutron cross sections of the FRENDY nuclear data processing system. The resonance upscattering effect is considered by (1) the variation of self-shielding factors (effective cross sections) due to the change in ultra-fine group spectrum and (2) the variation of group-to-group elastic scattering cross sections. In the verification calculations, impacts on the ultra-fine group spectrum, effective cross sections, and neutronics characteristics (the Doppler effect) are confirmed. The effect of energy group structure and the treatments of resonance upscattering on the Doppler effect through the variation of effective cross sections and the elastic scattering matrix are studied. The results indicate that the FRENDY can provide appropriate multi-group cross sections considering the resonance upscattering effect.

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 99 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.

Yamamoto, Akio*; Endo, Tomohiro*; Tada, Kenichi

Journal of Nuclear Science and Technology, 58(12), p.1343 - 1350, 2021/12

Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)An adaptive setting method of background cross sections is implemented to FRENDY/MG, which is a multi-group neutron cross section generation code. In the present adaptive setting method, the range of background cross section is initially divided into 10 equal intervals and unnecessary background cross section points, at which self-shielding factors or reaction rates can be accurately interpolated, are eliminated. If the interpolation accuracy in an interval is not sufficient, the interval is successively halved until sufficient interpolation accuracy is obtained. For accurate interpolation of self-shielding factor or reaction rates, the monotone cubic interpolation is used. Verification calculations are carried out for all isotopes in JENDL-4.0 and -4.0u. Calculation results indicate that typical numbers of background cross sections are from 10 to 25 when the monotone cubic interpolation and error tolerance of 0.01 for self-shielding factors are used.

Yamamoto, Akio*; Tada, Kenichi; Chiba, Go*; Endo, Tomohiro*

Journal of Nuclear Science and Technology, 58(11), p.1165 - 1183, 2021/11

Times Cited Count：7 Percentile：85.42(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, -54, 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.

Fujimoto, Nozomu*; Tada, Kenichi; Ho, H. Q.; Hamamoto, Shimpei; Nagasumi, Satoru; Ishitsuka, Etsuo

Annals of Nuclear Energy, 158, p.108270_1 - 108270_8, 2021/08

Times Cited Count：2 Percentile：36.99(Nuclear Science & Technology)Yamamoto, Akio*; Tada, Kenichi; Chiba, Go*; Endo, Tomohiro*

Transactions of the American Nuclear Society, 124(1), p.544 - 547, 2021/06

Verification calculations for the capability of multi-group cross section generation in FRENDY (FRENDY/MG) are carried out through the comparison of one-group reaction rates using the multi-group cross sections obtained by FRENDY/MG and NJOY2016. Three different neutron spectra (LWR, FR, and 1/E) are used to calculate one-group reaction rates. The discrepancies of one-group reaction rates are small for most cases, showing the validity of FRENDY/MG. The FRENDY/MG will be released as the part of FRENDY nuclear data processing system in the near future.

Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi; Endo, Tomohiro*

Transactions of the American Nuclear Society, 124(1), p.556 - 558, 2021/06

The FRENDY nuclear data processing code has been used to generate multi-group cross section libraries for the CBZ reactor physics code system. The newly generated libraries have been applied to neutronics calculations of a fast reactor core MET-1000, and several neutronics parameters are calculated. Calculations with other libraries generated by NJOY2016 have been also conducted, and differences in obtained neutronics parameters between the FRENDY-based library and the NJOY-based library have been quantified. Generally reasonable agreement between them has been obtained, so it has been demonstrated that the multi-group libraries for fast reactor neutronics calculations can be generated successfully by FRENDY. Detailed investigation on the impact of the difference in the processing codes on k-effective has been also carried out with a help of the perturbation theory, and the causes of the differences have been identified.

Endo, Tomohiro*; Noguchi, Akihiro*; Yamamoto, Akio*; Tada, Kenichi

Transactions of the American Nuclear Society, 124(1), p.184 - 187, 2021/06

This study confirmed that the sensitivity analysis of the alpha-eigenvalue can be carried even for non-neutron multiplication systems such as water-only systems. The preliminary results of nuclear data-induced uncertainties of alpha-eigenvalue were smaller than the differences between numerical and experimental results of alpha-eigenvalue. For further investigation, it is necessary to reconsider the experimental bias and the nuclear data-induced uncertainty in alpha-eigenvalue due to the thermal scattering law data of water.

Tada, Kenichi

Robutsuri No Kenkyu (Internet), (73), 5 Pages, 2021/03

This report is overview of FRENDY training course held in Oct. 2020. This report gives speaker's impressions.