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Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
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.
Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04
Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.
Doda, Norihiro; Uwaba, Tomoyuki; Ohgama, Kazuya; Yoshimura, Kazuo; Nemoto, Toshiyuki*; Tanaka, Masaaki; Yamano, Hidemasa
Nihon Kikai Gakkai Kanto Shibu Dai-29-Ki Sokai, Koenkai Koen Rombunshu (Internet), 5 Pages, 2023/03
An evaluation method for reactivity feedback due to core deformation during reactor power increase in sodium-cooled fast reactors is being developed for realistic core design evaluation. In this evaluation method, fuel assembly bowing was modeled with a beam element of the finite element method, and the assembly's pad contact between adjacent assemblies was modeled with a dedicated element which could consider the wrapper tube cross-sectional distortion and the pad stiffness depending on pad contact conditions. This fuel assembly bowing analysis model was verified for thermal bowing of a single assembly and assembly pad contact between adjacent assemblies in a core as past benchmark problems. The calculation results by this model showed good agreement with those of reference solutions of theoretical solutions or results by participating institutions in the benchmark. This study confirmed that the analysis model was able to calculate thermal assembly bowing appropriately.
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.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08
To confirm the applicability of the reactivity model, the authors have been conducting the benchmark exercises of the unprotected loss of heat sink event tests in a pool-type experimental fast reactor EBR-II. In the blind phase in the benchmark analyses using the plant dynamics analysis (1D) code in which the cold pool was modeled by means of the perfect mixing volume, it was found the increase of the core inlet temperature was evaluated lower than that of the measured data and the feedback reactivity was underestimated, because the thermal stratification in the cold pool was ignored. Then, the detailed model of the cold pool for the computational fluid dynamics (CFD) code was introduced and the 1D-CFD codes coupling method was applied to the benchmark analyses. It was confirmed that both the thermal stratification in the cold pool and the increase of the core inlet temperature were successfully reproduced.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Fujisaki, Tatsuya*; Murakami, Satoshi*; Vilim, R. B.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
In Japan Atomic Energy Agency, the multilevel simulation system which enables consistent evaluation from the whole plant behavior to the local phenomena is being developed to optimize plant design and enhance the safety of sodium-cooled fast reactors. To validate the coupling method in the MLS system, the 1D-CFD coupling method using Super-COPD for 1D plant dynamics analysis and Fluent for multi-dimensional CFD analysis was applied to the analyses of loss of flow tests in EBR-II. It was confirmed that it could predict multi-dimensional thermal-hydraulic phenomena such as thermal stratification in the upper plenum, Z-shaped pipe, and cold pool, holding the whole plant behavior simultaneously. Moreover, the applicability of the 1D-CFD coupling method to the evaluation of the phenomena in natural circulation conditions was confirmed by comparing the results of the 1D-CFD couple analyses and the measured data.
Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Yamano, Hidemasa; Igawa, Kenichi*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08
The numerical results of the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-302R and BOP-301) are discussed in order to validate the evaluation method of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD. By comparing the numerical results and the experimental data, the profiles of the increase of the core inlet temperature and the decrease of the reactor power calculated by Super-COPD were comparable with those of the experimental data and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS event.
Isobe, Mitsutaka*; Toi, Kazuo*; Matsushita, Hiroyuki*; Goto, Kazuyuki*; Suzuki, Chihiro*; Nagaoka, Kenichi*; Nakajima, Noriyoshi*; Yamamoto, Satoshi*; Murakami, Sadayoshi*; Shimizu, Akihiro*; et al.
Nuclear Fusion, 46(10), p.S918 - S925, 2006/10
Times Cited Count：30 Percentile：69.47(Physics, Fluids & Plasmas)no abstracts in English
Yoshimura, Kazuo; Aizawa, Kosuke; Ichikawa, Kenta; Mori, Takero; Yamada, Fumiaki
no journal, ,
no abstracts in English
Matsui, Kazuaki; Yoshimura, Kazuo; Aizawa, Kosuke; Ichikawa, Kenta; Yamada, Fumiaki
no journal, ,
no abstracts in English
Yoshimura, Kazuo; Ikeda, Makinori; Enuma, Yasuhiro; Aizawa, Kosuke
no journal, ,
JAEA has conducted a safety evaluation of impacts of double leakage at the PHTSs considering the passive safety features of Monju as the best-estimate evaluation for a DEC. The result shows that the total amount of leaked sodium can be reduced by the depressurization of the cover gas resulting from decrease in coolant inventory, i.e. negative pressure effects. The reactor coolant level required for decay heat removal, therefore, can be maintained even under double leakage at the PHTSs.
Mori, Takero; Sotsu, Masutake; Imaizumi, Yuya; Yoshimura, Kazuo; Fukano, Yoshitaka
no journal, ,
no abstracts in English
Nishino, Hiroyuki; Kurisaka, Kenichi; Yoshimura, Kazuo; Nishimura, Masahiro; Fukano, Yoshitaka
no journal, ,
no abstracts in English
Nishimura, Masahiro; Yoshimura, Kazuo; Taninaka, Hiroshi; Yamada, Fumiaki; Mori, Takero; Nishino, Hiroyuki; Fukano, Yoshitaka
no journal, ,
no abstracts in English
Yoshimura, Kazuo; Doda, Norihiro; Oki, Hiroshi*; Fujisaki, Tatsuya*; Tanaka, Masaaki; Vilim, R. B.*
no journal, ,
The CFD analysis of SHRT-17 test in the experimental fast bread reactor of EBR-II, was carried out to investigate the thermal stratification measured in the core upper plenum with the simulation result in the test. From the numerical simulation, it could be confirmed that the driving force of the natural circulation through the whole plant was influenced by the thermal stratification in the upper part of the core upper plenum and in the sloped part of the Z-shaped pipe.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Murakami, Satoshi*; Tanaka, Masaaki
no journal, ,
The multi-level simulation system with 1D-CFD coupling method which enables to evaluate various phenomena from the whole plant dynamics to the local thermal hydraulics has been developed. The numerical simulation of the ULOF test in the experimental fast reactor EBR-II in the U.S. is performed for validation study of the 1D-CFD coupling method, which combines a one-dimensional plant dynamics analysis (1D) code with a computational fluid dynamics (CFD) code. Through the numerical simulation, it was shown that the whole plant response and the multi-dimensional thermal hydraulics in the core upper plenum could be simulated. And the applicability of the 1D-CFD coupling method to plant scale analysis was confirmed in comparison with the experimental results.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
In the ULOHS tests performed in the experimental fast reactor U.S. EBR-II, the thermal stratification in the cold pool (CP) has influence on the whole plant behavior during the events because the secondary sodium pump tripped without scram nor tripping the primary pumps. In order to create the one-dimensional model for the CP of the plant dynamics analysis code, the multi-dimensional thermal hydraulics analyses using computational fluid dynamics (CFD) code were conducted to investigate the thermal hydraulics phenomena in the CP. It was found by comparison with the experimental data that the modeling of the detail sodium flow at the outlet of the intermediate heat exchanger, the leakage flow from the inner components to the cold pool, and the heat radiation from the CP to the atmosphere was important to the evaluation of the thermal stratification.
Yoshimura, Kazuo; Doda, Norihiro; Hamase, Erina; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
Sodium-cooled fast reactors have intrinsic safety features decreasing reactor power during the increase of the core inlet temperature by the feedback reactivity of the radial expansion of the core support plate. It is necessary for the composition of the core highly of secure to understand the influence of the safety features with high accuracy. In this paper, first, the 1D-CFD coupling method with cold pool as CFD region which enables the plant dynamics analyses taking account of the thermal stratification in cold pool was applied to the ULOHS (Unprotected Loss Of Heat Sink) test performed in the experimental fast reactor U.S. EBR-II and the evaluation of the core inlet temperature could be improved. Secondly, the sensitivity analyses concerning the core bowing reactivity were carried out with the aim of improving the evaluations of the core deformation reactivity and the applicability of the core bowing reactivity model to the test could be indicated.
