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Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Uwaba, Tomoyuki; Tanaka, Masaaki; Nemoto, Toshiyuki*
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 8 Pages, 2024/03
To investigate possibility of the insertion of the reactivity by the deflection of the upper core support plate, structural mechanics analyses of the domain consisting of the fuel assemblies and core support plates and evaluation of the reactivity due to the inclination of the fuel assemblies in EBR-II were carried out. As a result, it was indicated that the upper core support plate deflected downward larger at the low flowrate condition than that at the high flowrate condition and positive reactivity was inserted due to the inclination of the fuel assemblies at the low flowrate condition.
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 237(5), p.947 - 957, 2023/10
Times Cited Count：4 Percentile：69.72(Engineering, Multidisciplinary)Probabilistic risk assessment (PRA) is a method used to assess the risks associated with large and complex systems. However, the timing at which nuclear power plant structures, systems, and components are damaged is difficult to estimate if the risk of an external event is evaluated using conventional PRA based on event trees and fault trees. A methodology coupling thermal-hydraulic analysis with external event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID) is therefore proposed to overcome this limitation. A flood propagation model based on Bernoulli's theorem was applied to represent internal flooding in the turbine building of the pressurized water reactor. Uncertainties were also taken into account, including the flow rate of the floodwater source and the failure criteria for the mitigation systems. The simulated recovery actions included the operator isolating the floodwater source and using a drainage pump; these actions were modeled using several simplifications. Overall, the results indicate that combining isolation and drainage can reduce the conditional core damage probability upon the occurrence of flooding by approximately 90%.
Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.
OKondo, Yosuke*; Achouri, N. L.*; Al Falou, H.*; Atar, L.*; Aumann, T.*; Baba, Hidetada*; Boretzky, K.*; Caesar, C.*; Calvet, D.*; Chae, H.*; et al.
Nature, 620(7976), p.965 - 970, 2023/08
Times Cited Count：6 Percentile：92.64(Multidisciplinary Sciences)no abstracts in English
Ne at the transition into the island of inversion; Detailed structure study of NeWang, H.*; Yasuda, Masahiro*; Kondo, Yosuke*; Nakamura, Takashi*; Tostevin, J. A.*; Ogata, Kazuyuki*; Otsuka, Takaharu*; Poves, A.*; Shimizu, Noritaka*; Yoshida, Kazuki; et al.
Physics Letters B, 843, p.138038_1 - 138038_9, 2023/08
Times Cited Count：2 Percentile：68.16(Astronomy & Astrophysics)Detailed 
-ray spectroscopy of the exotic neon isotope Ne has been performed using the one-neutron removal reaction from Ne. Based on an analysis of parallel momentum distributions, a level scheme with spin-parity assignments has been constructed for Ne and the negative-parity states are identified for the first time. The measured partial cross sections and momentum distributions reveal a significant intruder p-wave strength providing evidence of the breakdown of the N = 20 and N = 28 shell gaps. Only a weak, possible f-wave strength was observed to bound final states. Large-scale shell-model calculations with different effective interactions do not reproduce the large p-wave and small f-wave strength observed experimentally, indicating an ongoing challenge for a complete theoretical description of the transition into the island of inversion along the Ne isotopic chain.
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.
Goux, P.*; Glessgen, F.*; Gazzola, E.*; Singh Reen, M.*; Focillon, W.*; Gonin, M.*; Tanaka, Tomoyuki*; Hagiwara, Kaito*; Ali, A.*; Sudo, Takashi*; et al.
Progress of Theoretical and Experimental Physics (Internet), 2023(6), p.063H01_1 - 063H01_15, 2023/06
Times Cited Count：0 Percentile：0.01(Physics, Multidisciplinary)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.
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.
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.
Doda, Norihiro; Uwaba, Tomoyuki; Yokoyama, Kenji; Nemoto, Toshiyuki*; Tanaka, Masaaki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 14 Pages, 2022/03
In sodium-cooled fast reactors, reactivity feedback is generated by thermal deformation of the core fuel assembly during core temperature rise. To utilize the core deformation reactivity as an inherent safety characteristic and to eliminate excessive conservativeness of core design in the safety evaluation, an evaluation method by coupling analyses of neutronics, thermal-hydraulics, and structural deformation has been developed. An experiment of unprotected loss-of-flow event in the experimental fast breeder reactor EBR-II was analyzed. The analysis results show that the core deformation reactivity has a negative feedback effect, and that the deformation reactivity is affected not only by the fuel movement but also by the movement of reflectors around the fuel. As a result, the availability of the evaluation method for core deformation reactivity feedback by coupled analysis approach is confirmed.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Times Cited Count：3 Percentile：35.51(Nuclear Science & Technology)Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
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.
Nakasone, Shunya*; Yokoyama, Sumi*; Takahashi, Tomoyuki*; Ota, Masakazu; Kakiuchi, Hideki*; Sugihara, Shinji*; Hirao, Shigekazu*; Momoshima, Noriyuki*; Tamari, Toshiya*; Shima, Nagayoshi*; et al.
Plasma and Fusion Research (Internet), 16, p.2405035_1 - 2405035_5, 2021/02
Removal of impurities such as organic and other types of dissolved matters from environmental water samples is required for precise analysis of tritium with a liquid scintillation counting method. In general, a distillation method is a conventional one for tritium analysis in environmental water samples, but is a time-consuming process that takes 24 hours for removal of impurities. We have proposed a rapid pretreatment method for tritium analysis, that uses ion exchange resins. In this study, we performed batch experiments, to evaluate the effectiveness of the ion exchange resins on the tritium measurement. The results obtained demonstrated that removal of impurities in the sample water by ion exchange resins can be achieved during a short period of time (i.e., in 5 min).
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11
Probabilistic risk assessment (PRA) is one of the methods used to assess the risks associated with large and complex systems. When the risk of an external event is evaluated using conventional PRA, a particular limitation is the difficulty in considering the timing at which nuclear power plant structures, systems, and components fail. To overcome this limitation, we coupled thermal-hydraulic and external-event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID). Internal flooding was chosen as the representative external event, and a pressurized water reactor plant model was used. Equations based on Bernoulli's theorem were applied to flooding propagation in the turbine building. In the analysis, uncertainties were taken into account, including the flow rate of the flood water source and the failure criteria for the mitigation systems. In terms of recovery action, isolation of the flood water source by the operator and drainage using a pump were modeled based on several assumptions. The results indicate that the isolation action became more effective when combined with drainage.
Tanaka, Yoichi; Tamaki, Hitoshi; Zheng, X.; Sugiyama, Tomoyuki
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.308 - 315, 2020/10
Dynamic probabilistic risk assessment (PRA) is a method for improving the realism and completeness of conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a simplified accident sequence and compared the results for each method. Quasi-Monte Carlo sampling was found to be the most effective method in this case.
CsCl-supplemented diet to evaluate the effects of low dose-rate exposureTanaka, Sota; Kinouchi, Tadatoshi*; Fujii, Tsuguru*; Imanaka, Tetsuji*; Takahashi, Tomoyuki*; Fukutani, Satoshi*; Maki, Daisuke*; Notomi, Akihiro*; Takahashi, Sentaro*
Scientific Reports (Internet), 10, p.16055_1 - 16055_7, 2020/09
Times Cited Count：7 Percentile：49.75(Multidisciplinary Sciences)Since the Fukushima Dai-ichi Nuclear Power Plant (FDNPP) accident, morphological abnormalities of the lepidopteran insects have been reported. However, it is unclear whether the abnormalities were caused directly by radiation because they did not study on absorbed dose and dose-effect relationship. In this study, we conducted an internal exposure experiment on silkworm using CsCl-supplemented artificial diet and estimated the absorbed dose to evaluate the morphological abnormalities in silkworm. The ratio of wing to whole body of pupae was compared between the CsCl-exposured and control groups and no significant differences were observed between the groups. This result suggest that morphological abnormalities in lepidopterans are unlikely due to direct radiation effects from Cs contamination after the FDNPP accident.
Nakasone, Shunya*; Yokoyama, Sumi*; Takahashi, Tomoyuki*; Ota, Masakazu; Kakiuchi, Hideki*; Sugihara, Shinji*; Hirao, Shigekazu*; Momoshima, Noriyuki*; Tamari, Toshiya*; Shima, Nagayoshi*; et al.
Plasma and Fusion Research (Internet), 15, p.2405027_1 - 2405027_3, 2020/05
A quick preprocessing system for tritium analysis of environmental samples is important to judge environmental influence of tritium releases due to accident or tritium-handling facilities. Analysis of tritium in water samples with liquid scintillation counting method requires removal of impurities such as organic matter and ion species from water samples. Generally, a distillation method is adopted as a pretreatment of analysis for tritium; however, the distillation method is a time-consuming process. The aim of this study is to evaluate a rapid pretreatment method for tritium analysis with ion exchange resin. From batch and column experiments that used inland water and ion exchange resin, we confirmed removals of impurities of the water sample and that the removal of impurities was possible for a short time (by 5 minutes).
