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Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 60(11), p.1372 - 1385, 2023/11
Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10
The latest Japanese nuclear data library, JENDL-5, was released in December 2021. In JENDL-5, nuclear reaction cross sections for Gd-155 and Gd-157 were modified in addition to many heavy nuclides such as U-235. Fission yields and decay data, which are essential to characterize burnup fuels, were completely revised. This study investigated the effects of the nuclear data revisions from JENDL-4.0 to JENDL-5 on the neutronic characteristics of burnup fuels to validate JENDL-5. Burnup calculations of the 9x9 STEP-3 BWR fuel assembly based on the OECD/NEA Phase III-C benchmark were performed using JENDL-4.0 and JENDL-5. As a result, the k
for JENDL-5 was smaller than that of JENDL-4.0 throughout the burnup, with a large difference of about 600 pcm at 12 GWd/t, around the peak of the k. Above 20 GWd/t, the difference in k increases with increasing burnup value, reaching nearly 600 pcm at 50 GWd/t. In addition, this study investigates which nuclear data contribute significantly to the difference in k by performing burnup calculations with replacing nuclear data of individual nuclides from JENDL-4.0 to JENDL-5.
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.
Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*
EPJ Web of Conferences, 281, p.00008_1 - 00008_9, 2023/03
The applicability of Akaike's Bayesian Information Criterion (ABIC) to covariance modeling in the cross-section adjustment method was investigated. One of the most important things for a reliable cross-section adjustment method is giving a suitable covariance matrix. However, since we cannot know the true covariance matrix in advance, we usually estimate and assume it. To judge the goodness of the covariance matrix modeling, a metric is desirable. As a candidate for this metric, we focus on ABIC which is one of the information criteria in Bayesian inference, because the cross-section adjustment method is often discussed within the framework of Bayesian inference. In the conventional cross-section adjustment method, incorporation of the analysis model uncertainty in a covariance matrix still requires ad hoc treatment. In JAEA, the integral experimental database for fast reactors has been developed and the adjusted cross-section set ADJ2017 has been created based on this database. Many of the core characteristics in the database have been analyzed by a deterministic method. Therefore, the predicted core characteristics have non-negligible uncertainties with correlations due to some numerical approximations. However, the evaluations of the uncertainties and their correlations are still challenging issues. In addition, there would be unknown uncertainties that experimenters and analysts of reactor physics experiments could not recognize. To judge the goodness of the covariance matrix related to these uncertainties, the applicability of ABIC to the cross-section adjustment method was investigated.
Fukui, Yuhei*; Endo, Tomohiro*; Yamamoto, Akio*; Maruyama, Shuhei
EPJ Web of Conferences, 281, p.00006_1 - 00006_9, 2023/03
We developed a new nuclear data adjustment method for experimental data containing outliers. This method mitigates the effect of outliers by applying M-estimation, a type of robust estimation, to the conventional nuclear data adjustment method using sensitivity coefficients. Based on the M-estimation, we derived a weighted nuclear data adjustment formula and developed a weight calculation method. The weighted nuclear data adjustment formula was derived by weighting the function to take the extremum of the conventional nuclear data adjustment. The weighting of each nuclear characteristic is calculated from the difference between the measured and calculated values of the nuclear characteristic. This weight calculation method can evaluate the validity of each nuclear characteristic by considering correlations between nuclear characteristics using singular value decomposition. The proposed method and the conventional method were compared and verified by twin experiments. In the twin experiments, the nuclear data were adjusted using experimental data that intentionally included outliers. As a result of twin experiments, it was confirmed that the nuclear data were adjusted robustly and appropriately even with the experimental data containing outliers.
Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Nagaya, Yasunobu
JAEA-Data/Code 2022-009, 208 Pages, 2023/02
JAEA-Data-Code-2022-009.pdf:3.87MB
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.
Watanabe, Tomoaki; Tada, Kenichi; Endo, Tomohiro*; Yamamoto, Akio*
Journal of Nuclear Science and Technology, 11 Pages, 2023/00
Times Cited Count：0 Percentile：82.84(Nuclear Science & Technology)The burnup calculations for estimating the nuclide composition of the spent fuel are highly dependent on nuclear data. Many nuclides in the latest version of the Japanese Evaluated Nuclear Data Library JENDL-5 were modified from JENDL-4.0 and the modification affects the burnup calculations. This study confirmed the validity of JENDL-5 in the burnup calculations. The PIE data of Takahama-3 was used for the validation. The effect of modifications of the parameters, e.g., cross sections and fission yields, from JENDL-4.0 to JENDL-5 on the nuclide compositions was quantitatively investigated. The calculation results showed that JENDL-5 has a similar performance to JENDL-4.0. The calculation results also revealed that the modifications of the cross sections of actinide nuclides, fission yields, and thermal scattering low data of hydrogen in H
O affected the nuclide compositions of PWR spent fuels.
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.
Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro; Pyeon, C. H.*; Yamamoto, Akio*; Endo, Tomohiro*
Nuclear Science and Engineering, 20 Pages, 2023/00
In this study, we have demonstrated that data assimilation using lead and bismuth sample reactivities measured in the Kyoto University Critical Assembly A-core can successfully reduce the uncertainty of the coolant void reactivity in accelerator-driven systems derived from inelastic-scattering cross-sections of lead and bismuth. We re-evaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the data assimilation formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties, and performed data assimilation using the reactor analysis code system MARBLE. The high-sensitivity coefficients of the sample reactivities to lead and bismuth allowed us to reduce the cross-section-induced uncertainty of the void reactivity of the accelerator-driven system from 6.3% to 4.8%, achieving a provisional target accuracy of 5% in this study. Furthermore, we demonstrated that the uncertainties arising from other dominant factors, such as minor actinides and steel, can be effectively reduced by using integral experimental data sets for the unified cross-section dataset ADJ2017.
Fujita, Natsuko; Miyake, Masayasu; Matsubara, Akihiro*; Ishii, Masahiro*; Watanabe, Takahiro; Jinno, Satoshi; Nishio, Tomohiro*; Ogawa, Yumi; Yamamoto, Yusuke; Kimura, Kenji; et al.
Dai-23-Kai AMS Shimpojiumu Hokokushu, p.1 - 4, 2022/12
The JAEA-AMS-TONO facility at the Tono Geoscience Center, JAEA has three accelerator mass spectrometers. We report the present status of the JAEA-AMS-TONO.
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.
Katano, Ryota; Yamamoto, Akio*; Endo, Tomohiro*
Nuclear Science and Engineering, 196(10), p.1194 - 1208, 2022/10
Times Cited Count：0 Percentile：40.11(Nuclear Science & Technology)In this study, we propose the ROM-Lasso method that enables efficient evaluation of sensitivity coefficients of neutronics parameters to cross-sections. In the proposed method, a vector of sensitivity coefficients is expanded by subspace bases, so-called Active Subspace (AS) based on the idea of Reduced Order Modeling (ROM). Then, the expansion coefficients are evaluated by the Lasso linear regression between cross-sections and neutronics parameters obtained by the random sampling. The proposed method can be applied in the case where the adjoint method is difficult to be applied since the proposed method uses only forward calculations. In addition, AS is an effective subspace that can expand the vector of sensitivity coefficients with the lower number of dimension. Thus, the number of unknows is reduced from the original number of input parameters and the calculation cost is dramatically improved compared to the Lasso regression without AS. In this paper, we conducted ADS burnup calculations as a verification. We have shown how AS bases are obtained and the applicability of the proposed method.
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 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.
Katano, Ryota; Yamamoto, Akio*; Endo, Tomohiro*
Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), p.2032 - 2041, 2022/05
We have proposed the ROM-Lasso method to perform an efficient evaluation of the sensitivity coefficients of ADS core parameters to cross sections without major modification of the core analysis system. In the ROM-Lasso method, the sensitivity coefficient vector is expanded via the subspace bases so-called Active Subspace (AS), and the effective number of unknowns is reduced. Then, the expansion coefficients are determined via the penalized linear regression with the core parameters obtained by the random sampling, and the sensitivity coefficient vector is estimated. Owing to the AS, the required number of the core calculations is dramatically reduced in the ROM-Lasso method. In this work, we take the sensitivity coefficient evaluation of the coolant void reactivity at the end of the cycle for example and demonstrate how estimation accuracy depends on the number of samples and the AS.
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, -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.
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.
