Yamamoto, Akio*; Endo, Tomohiro*; Chiba, Go*; Tada, Kenichi
Nuclear Science and Engineering, 196(11), p.1267 - 1279, 2022/11
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.
Konno, Chikara; Tada, Kenichi; Kwon, Saerom*
Proceedings of 14th International Conference on Radiation Shielding and 21st Topical Meeting of the Radiation Protection and Shielding Division (ICRS-14/RPSD 2022) (Internet), p.440 - 443, 2022/11
Neutron spectra inside a sphere of 1 m in radius, made of one natural isotope with unresolved resonance data, with an isotropic neutron source of 20 MeV at the center were calculated with the ANISN code and JENDL-4.0 MATXS file MATXSLIB-J40. Then unphysical neutron spectra produced in unresolved resonance data processing with the NJOY code were obtained. We examined its reasons and specified that unrealistic cross sections in dips between resonances caused the unphysical neutron spectra. We also demonstrated that this problem was solved by modifying NJOY.
Kamiya, Tomohiro; Ono, Ayako; Tada, Kenichi; Akie, Hiroshi; Nagaya, Yasunobu; Yoshida, Hiroyuki; Kawanishi, Tomohiro
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/11
JAEA started to develop the advanced reactor analysis code JAMPAN (JAEA advanced multi-physics analysis platform for nuclear systems). The current version of JAMPAN handles the continuous energy Monte Carlo code MVP and the detailed thermal-hydraulics analysis code for multiphase and multicomponent JUPITER. JAMPAN is designed to consider the extensibility and it does not depend on the analysis codes. All calculations in JAMAPAN are not directly connected. JAMPAN has data containers, and all input and output data of each analysis code are set in these data containers. JAMPAN will easily exchange the calculation codes and add the other calculations, e.g., structure calculation and irradiation calculation since the input and the output format of each code has no impact on the other calculation codes. The 4 by 4 pin-cell geometry was used as the demonstration calculation of JAMPAN and the physically reasonable calculation results were obtained.
Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi
Journal of Nuclear Science and Technology, 8 Pages, 2022/06
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; 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.
Tada, Kenichi; Kondo, Ryoichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 8 Pages, 2022/00
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.
Yamamoto, Akio*; Endo, Tomohiro*; Tada, Kenichi
Journal of Nuclear Science and Technology, 58(12), p.1343 - 1350, 2021/12
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
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
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.
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.
JAEA-Data/Code 2020-014, 30 Pages, 2020/10
The decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. In the process of the decommissioning, preventing radiation exposure of workers is imperative originating in nuclear criticality of fuel debris. This study provides the handy tool enabling the analysis on nuclear criticality of fuel debris. The developed analysis tool named as HAND enables estimation of the criticality of fuel debris in short time. HAND deduces the range of parameters such as the size and composition, in which the criticality of fuel debris is specified. By selecting the range of the parameters using HAND in advance, total calculation time of the detail analysis will be reduced. Since the input data of HAND is designed to be simple and the output of HAND is to be user friendly, this tool is expected to be also an intuitive tool to study the criticality of fuel debris. This report explains the outline of the HAND and input instructions for HAND.
Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Tada, Kenichi
Kaku Deta Nyusu (Internet), (127), p.1 - 10, 2020/10
The 32nd annual meeting and the subgroup meeting of the Working Party on International Nuclear Data Evaluation Co-operation (WPEC) under the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) was held via a web meeting system from 11 to 15 in May in 2020. The activities about nuclear data measurement and evaluation of each region or country were reported at the annual meeting, and the SG activities were discussed at the subgroup meetings. The summary of these meetings are reported.
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
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.
Yamamoto, Akio*; Endo, Tomohiro*; Tada, Kenichi
Transactions of the American Nuclear Society, 122(1), p.714 - 717, 2020/06
A generation capability of multi-group cross sections from point-wise cross sections in ACE files is being developed as a function of the nuclear data processing code FRENDY. This presentation describes features of this function and comparison of the processing results between this function and GROUPR module in NJOY.
Proceedings of International Conference on the Physics of Reactors; Transition To A Scalable Nuclear Future (PHYSOR 2020) (USB Flash Drive), 8 Pages, 2020/03
The probability table is widely used for continuous energy Monte Carlo calculation codes to treat the self-shielding effect in the unresolved resonance region. The ladder method is used to calculate the probability table. This method generates a lot of pseudo resonance structures using random numbers based on the averaged resonance parameters. The probability table affects the ladder number. i.e., number of pseudo resonance structures. The ladder number has large impact on the generation time of the cross section library. In this study, the appropriate ladder number is investigated. The probability table of all nuclides prepared in JENDL-4.0 is used to investigate the appropriate ladder number. The comparison results indicate that the differences of the probability table are enough small when the ladder number is 100.
Tada, Kenichi; Kunieda, Satoshi
KURNS-EKR-5, p.229 - 232, 2019/12
The R-matrix limited formula is formatted by the current nuclear data format and it is adopted some nuclei in the latest evaluated nuclear data library. Since the processing of the R-matrix limited formula is significantly different to the other resonance formulae, it is difficult to treat this formula without large modification of the nuclear data processing code. In this study, we implemented one of the Rmatrix code AMUR to treat this formula in FRENDY. The processing results of FRENDY are compared to those of NJOY2016 to verify FRENDY. The comparison results indicate that FRENDY appropriately treat the R-matrix limited formula with similar computational time.