Development of a design optimization framework for sodium-cooled fast reactors, 2; Development of optimization analysis control function

Doda, Norihiro; Nakamine, Yoshiaki*; Kuwagaki, Kazuki; Hamase, Erina; Kikuchi, Norihiro; Yoshimura, Kazuo; Matsushita, Kentaro; Tanaka, Masaaki

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 28, 5 Pages, 2023/05

As a part of the development of the "Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to automatically optimize the life cycle of innovative nuclear reactors including fast reactors, ARKADIA-design is being developed to support the optimization of fast reactor in the conceptual design stage. ARKADIA-Design consists of three systems (Virtual plant Life System (VLS), Evaluation assistance and Application System (EAS), and Knowledge Management System (KMS)). A design optimization framework controls the connection between the three systems through the interfaces in each system. This paper reports on the development of an optimization analysis control function that performs design optimization analysis combining plant behavior analysis by VLS and optimization study by EAS.

Development of ARKADIA-Design for design optimization support; Application of coupling method using multi-level simulation technique for plant thermal-hydraulics analysis

Doda, Norihiro; Yoshimura, Kazuo; Hamase, Erina; Yokoyama, Kenji; Uwaba, Tomoyuki; Tanaka, Masaaki

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09

ARKADIA-Design is being developed to support the optimization of sodium-cooled fast reactors in the conceptual design stage. Design optimization requires various types of numerical analysis: 1-D plant dynamics analysis for efficient evaluation of various design options and multi-dimensional analysis for a detailed evaluation of local phenomena, including multi-physics. For those analyses, ARKADIA-Design performs whole plant analyses based on the multi-level simulation (MLS) technique in which the analysis codes are coupled to simulate the phenomena in an intended degree of resolution. This paper describes an outline of the coupling analysis methods in the MLS of the ARKADIA-Design and the numerical simulations of the experimental fast breeder reactor EBR-II tests by the coupled analysis.

Damage evaluations for BWR lower head in severe accident based on multi-physics simulations

Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Furuta, Takuya; Kaji, Yoshiyuki

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07

Development of multi-level simulation system for core thermal-hydraulics coupled with plant dynamics analysis; Prediction of transient temperature distribution in a subassembly under inter-subassembly heat transfer effect

Doda, Norihiro; Hamase, Erina; Kikuchi, Norihiro; Tanaka, Masaaki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

In conventional design studies of sodium-cooled fast reactors, plant dynamics and local phenomena were evaluated separately by using simple models and detailed models, respectively, and their interaction was considered through the boundary conditions settings with conservativeness for each individual analysis. Thus, the final result through the analyses may contain excessive conservativeness. Therefore, JAEA began to develop a multi-level simulation system in which detailed analysis codes are coupled with a plant dynamics analysis code. Focusing on core thermal-hydraulics, a coupled analysis method using a plant dynamics analysis code Super-COPD and a subchannel analysis code ASFRE has been developed. The analysis on a test in the experimental fast reactor EBR-II was performed to validate the coupled analysis. Through the comparison of the analysis results and the measurement, it was confirmed that the coupled analysis could predict the transient temperature distribution in the subassembly, and the multi-level simulation by changing the level of detail in analysis model could be performed for core thermal-hydraulics.

Development of numerical analysis codes for multi-level and multi-physics approaches in an advanced reactor design study

Tanaka, Masaaki; Doda, Norihiro; Mori, Takero; Yokoyama, Kenji; Uwaba, Tomoyuki; Okajima, Satoshi; Matsushita, Kentaro; Hashidate, Ryuta; Yada, Hiroki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Japan Atomic Energy Agency is developing an innovative design system named ARKADIA to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In the first phase of its development, ARKADIA-Design for design study and ARKADIA-Safety for safety assessment will be developed individually. In this paper, focusing on the ARKADIA-Design, the concept of the system is described and numerical analysis codes to be used for the multi-level and multi-physics analyses are introduced. Descriptions of the practical functions composed by the analysis codes and the representative problems in application studies for validation are introduced.

Development of neutronics, thermal-hydraulics, and structure mechanics coupled analysis method on integrated numerical analysis for design optimization support in fast reactor

Doda, Norihiro; Uwaba, Tomoyuki; Nemoto, Toshiyuki*; Yokoyama, Kenji; Tanaka, Masaaki

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 26, 4 Pages, 2021/05

For design optimization of fast reactors, in order to consider the feedback reactivity due to thermal deformation of the core when the core temperature rises, which could not be considered in the conventional design analysis, a neutronics, thermal-hydraulics, and structure mechanics coupled analysis method has been developed. Neutronics code, plant dynamics code, and structural mechanics code are coupled by a control module in python script. This paper outlines the coupling method of analysis codes and the results of its application to an experiment in an actual plant.

Development of neutronics and thermal-hydraulics coupled analysis method on platform for design optimization in fast reactor

Doda, Norihiro; Hamase, Erina; Yokoyama, Kenji; Tanaka, Masaaki

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 25, 4 Pages, 2020/06

With the aim of advancing the design optimization in fast reactors, neutronics and thermal-hydraulics coupled analysis method which can consider the temporal change of neutron flux distribution in the core has been developed. A three-dimensional neutronics analysis code and a plant dynamics analysis code are coupled on a platform using Python programing. In this report, outlines of the coupling method of analysis codes, the results of its application to the actual plant under a virtual accidental condition, and the future development is described.

Conceptual design study of beam window for accelerator-driven system with subcriticality adjustment rod

Sugawara, Takanori; Eguchi, Yuta; Obayashi, Hironari; Iwamoto, Hiroki; Tsujimoto, Kazufumi

Nuclear Engineering and Design, 331, p.11 - 23, 2018/05

https://doi.org/10.1016/j.nucengdes.2018.02.011

This study aims to perform the coupled analysis for the feasible beam window concept. To mitigate the design condition, namely to reduce the necessary proton beam current, subcriticality adjustment rod (SAR) was installed to the ADS core. The burnup analysis was performed for the ADS core with SAR and the results indicated that the maximum proton beam current during the burnup cycle was reduced from 20 to 13.5 mA. Based on the burnup analysis result, the coupled analysis; particle transport, thermal hydraulics and structural analyses, was performed. As the final result, the most robust beam window design; the hemisphere shape, the outer radius = 235 mm, the thickness at the top of the beam window = 3.5 mm and the factor of safety for the buckling = 9.0, was presented. The buckling pressure was 2.2 times larger than the previous one and more feasible beam window concept was presented through this study.

Fundamental validation of fluid-structure thermal interaction simulation code for thermal striping in sodium-cooled fast reactors with parallel triple jets mixing experiments

Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*

Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-6) (Internet), 12 Pages, 2016/09

A numerical simulation code named MUGTHES which can deal with conjugate heat transfer problem between the fluid and the structure parts has been developed in order to predict the thermal response in the structure for estimation of the thermal fatigue issue. To perform fundamental validation of the MUGTHES, the benchmark simulation was considered using the experiment of planar triple parallel jets mixing sodium test (PLAJEST). Since it was known by literatures that three representative flow mixing patterns were shown in accordance with the velocity rate of the side jets to the center jet, three typical experimental conditions in the PLAJEST were employed as boundary conditions for the benchmark. Through the numerical simulations, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model to simulate thermal striping phenomena was confirmed.

Development of coupled method of fluid-structure thermal interaction simulation and thermal stress analysis for T-junction piping system

Tanaka, Masaaki; Miyake, Yasuhiro*; Karakida, Yasuhisa*

Proceedings of 2nd International Conference on Maintenance Science and Technology (ICMST-Kobe 2014), p.79 - 80, 2014/11

A coupled method of fluid-structure thermal interaction simulation and thermal stress analysis has been developed through the interface program to carry out direct numerical estimation of the thermal fatigue. The prototype method was applied to the thermal mixing phenomena in T-junction Piping System.