Fast burnup calculation method based on neutron spectrum reconstruction with proper orthogonal decomposition and regression model

Watanabe, Tomoaki; Aizawa, Naoto*; Chiba, Go*; Tada, Kenichi; Yamamoto, Akio*

Proceedings of International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025) (Internet), p.288 - 297, 2025/04

Currently, a major burnup calculation method for the nuclide composition of nuclear fuel conducts neutron transport calculations at each burnup step to account for changes in the neutron spectrum. While this method is highly accurate, the large computational cost of neutron transport calculations can be problematic. Therefore, a fast burnup calculation method based on neutron spectrum reconstruction with the proper orthogonal decomposition (POD) and regression model is investigated. In this method, dimensionality reduction by POD is applied to many neutron fluxes obtained from detailed burnup calculations for various input parameter sets, and regression models are constructed to connect the dimensionality-reduced neutron fluxes and parameters. By substituting arbitrary input parameters to the regression models, the neutron flux is reconstructed and the burnup calculation is performed. This method performs burnup calculations that consider changes in the neutron spectrum based on input conditions without neutron transport calculations. The present method was applied to a PWR UO fuel pin cell model. The results show the nuclide inventory can be calculated with a prediction accuracy within a few percent. In addition, it is found that the calculation error is dominated by the regression models, which implies the further improvement of the regression models leads to improving the accuracy.

Development of nuclear data processing code FRENDY version 2

Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Konno, Chikara; Kondo, Ryoichi; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*

Journal of Nuclear Science and Technology, 61(6), p.830 - 839, 2024/06

https://doi.org/10.1080/00223131.2023.2278600

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.

ACE-FRENDY-CBZ; A New neutronics analysis sequence using multi-group neutron transport calculations

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

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

https://doi.org/10.1080/00223131.2022.2087783

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.

Si-addition contributes to overcoming the strength-ductility trade-off in high-entropy alloys

Wei, D.*; Gong, W.; Tsuru, Tomohito; Lobzenko, I.; Li, X.*; Harjo, S.; Kawasaki, Takuro; Do, H.-S.*; Bae, J. W.*; Wagner, C.*; et al.

International Journal of Plasticity, 159, p.103443_1 - 103443_18, 2022/12

https://doi.org/10.1016/j.ijplas.2022.103443

Development of nuclear data processing code FRENDY version 2

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.

Verification of the multi-group generation capability of FRENDY nuclear data processing code for recent nuclear data through comparison of one-group reaction rates

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.

Development of a reference database for beta-delayed neutron emission

Dimitriou, P.*; Dillmann, I.*; Singh, B.*; Piksaikin, V.*; Rykaczewski, K. P.*; Tain, J. L.*; Algora, A.*; Banerjee, K.*; Borzov, I. N.*; Cano-Ott, D.*; et al.

Nuclear Data Sheets, 173, p.144 - 238, 2021/03

https://doi.org/10.1016/j.nds.2021.04.006

-delayed neutron emission has been of interest since the discovery of nuclear fission. In nuclear power reactors, delayed-neutron data play a crucial role in reactor kinetics calculations and safe operation. -delayed neutron data also have a significant impact in the field of nuclear structure and astrophysics especially as nuclei farther away from stability are explored at the new generation of radioactive beam facilities. Several compilations of -decay half-lives and delayed-neutron emission probabilities are available, however, complete documentation of measurements and evaluation procedures is often missing for these properties. Efforts to address this gap in nuclear data and create an updated compilation and evaluation of -delayed neutron properties were undertaken under the auspices of the International Atomic Energy Agency (IAEA) which formed a Coordinated Research Project (CRP) on "Development of a Reference Database of Beta-delayed Neutron Emission Data". In this paper we summarize the work that was performed and present the results of the CRP.

Measurement of MA reaction rates under sub-critical condition with spallation neutron source in A-core of KUCA for ADS

Oizumi, Akito; Fukushima, Masahiro; Tsujimoto, Kazufumi; Chiba, Go*; Yamanaka, Masao*; Sano, Tadafumi*; Pyeon, C. H.*

KURNS Progress Report 2018, P. 38, 2019/08

In the nuclear transmutation system such as ADS, the nuclear data validation of MA is required to reduce the uncertainty caused by the nuclear data of MA. This study aims to measure the fission reaction rate ratios (FRRs) of Neptunium-237 (Np) or Americium-241 (Am) to Uranium-235 (U) by using a back-to-back (BTB) fission chamber in the KUCA built as a sub-critical core (k = 0.998) with the nuclear spallation neutron source. The result showed that the measured FRRs of Np/U and Am/U were 0.014 0.002 and 0.023 0.005, respectively. These measured values will be used for verification of evaluated nuclear data by conducting detailed analyses.

Research and development roadmap for reactor physics 2017; Future of reactor physics projected by next generations

Yamamoto, Akio*; Chiba, Go*; Kirimura, Kazuki*; Miki, Yosuke*; Yokoyama, Kenji

Nihon Genshiryoku Gakkai-Shi ATOMO, 60(4), p.241 - 245, 2018/04

no abstracts in English

Cutting-edge studies on nuclear data for continuous and emerging need, 1; Diversifying nuclear applications and need for the nuclear data

Suyama, Kenya; Kunieda, Satoshi; Fukahori, Tokio; Chiba, Go*

Nihon Genshiryoku Gakkai-Shi ATOMO, 59(10), p.598 - 602, 2017/10

The nuclear data is the data on the reaction probability between the neutron and the nuclide in a narrow sense. However generally speaking, it is the data describing the physical change of the nuclide and the status of the nuclear ration. Since Japan had started the nuclear energy development, the nuclear data has been one of the most important technical development theme. Now, the nuclear data library of Japan, i.e., JENDL, is well recognized internationally because of the highest-accuracy and fully-furnished types of the included data. This serial lecture describes the significance and the status of the nuclear data development, the international trend, and the direction of the future development.