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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.

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 9 Pages, 2022/05

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.

Maruyama, Shuhei

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

This paper proposes a new homogenization method, "Boundary Condition Free Homogenization (BCFH)". The traditional homogenization method separates the core calculation and the cell (assembly) calculation by assuming a specific boundary condition or a peripheral region in the cell calculation. Nevertheless, there are ambiguities and approximation in these assumptions, and they can also cause a decline in accuracy. BCFH aims to avoid these problems and improve the accuracy in the cell calculation such as homogenization. We imposed the conditions that the physical quantities in the cell related to the reaction rate preservation is preserved for any incoming partial current, during the homogenization. That is, the response matrices of cell average (or total) flux and outgoing partial current, to be the same form between heterogeneous and homogeneous system. As a result, homogenized parameters, such as cross-sections, superhomgenization factors, and discontinuity factors, are no longer dependent on a specific boundary condition. The new homogenized parameters obtained in this way are extended from the conventional vector form to the matrix form in BCFH. To investigate the performance of BCFH, numerical tests are done for the simplified models which originates in 750MW-class sodium-cooled fast reactor with MOX fuel core in Japan. It is found that BCFH is particularly effective in evaluating control rod reactivity worth and reaction rate distribution, compared to the traditional method. We conclude that the BCFH can be a promising homogenization concept for core neutronic analysis.

Kuwagaki, Kazuki; Yokoyama, Kenji

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

In Japan Atomic Energy Agency (JAEA), an innovative design approach named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) for the advanced nuclear reactors is currently under development. One of the tasks in ARKADIA is to build a system that automatically optimizes core and fuel designs by conducting core neutronic and thermal-hydraulic calculations, fuel integrity evaluations, and plant dynamic analyses. This system will be implemented to automatically find an optimal design that minimizes (or maximizes) objective function defined by a core performance, while varying the core and fuel design parameters such as fuel pin diameter, core height and diameter. In this study, as the first step of the system development, we focused only on core neutronic design and conducted a study of automatic optimization. As the optimization algorithm, Bayesian optimization (BO), which is an effective method for optimization problems with expensive computation cost of objective function, was selected. The applicability of BO was studied based on single- and two-objective optimization examples of core neutronic design in a toy model. As a result, in the former, it was confirmed that BO can obtain the optimal solution, which well matches the reference solution calculated by a brute force calculation, with a small number of required calculation executions. Its usability on core neutronic designs, where the computation cost per case is large, was confirmed. In the latter, it was shown that BO can obtain a pareto solutions-set that shows good agreement with the reference solution.

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.

Gunji, Satoshi; Araki, Shohei; Suyama, Kenya; Izawa, Kazuhiko

The fuel debris is expected to have not only heterogeneous but also non-uniform compositions. Therefore, the calculation method used in their criticality management is required to be validated experimentally. In this study, several core configurations of a new critical assembly "STACY" of JAEA with non-uniform arrangements of uranium oxide fuel rods, concrete rods and stainless steel rods, which are components of the fuel debris, were studied. In each case, the median value of 100 sample patterns was larger than the mean effective multiplication factor. It was also confirmed that there are differences in the effective multiplication factor of more than one dollar by the pattern changing, and that the neutron spectra can change significantly by changing the local neutron moderation conditions. In particular, the effective multiplication factor became smaller when over-moderated regions with large water-to-fuel ratios were formed in the core configurations due to increases in thermal neutron absorption. Such criticality experiments with non-uniform arrangements of multiple compositions will be useful to evaluate the validity of the calculation code.