Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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%.
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 60(4), p.359 - 373, 2023/04
Times Cited Count：5 Percentile：78.52(Nuclear Science & Technology)Probabilistic risk assessment (PRA) is an essential approach to improving the safety of nuclear power plants. However, this method includes certain difficulties, such as modeling of combinations of multiple hazards. Seismic-induced flooding scenario includes several core damage sequences, i.e., core damage caused by earthquake, flooding, and combination of earthquake and flooding. The flooding fragility is time-dependent as the flooding water propagates from the water source such as a tank to compartments. Therefore, dynamic PRA should be used to perform a realistic risk analysis and quantification. This study analyzed the risk of seismic-induced flooding events by coupling seismic, flooding, and thermal-hydraulics simulations, considering the dependency between multiple hazards explicitly. For requirements of safety improvement, especially in light of the Fukushima Daiichi Nuclear Power Plant accident, sensitivity analysis was performed on the seismic capacity of systems, and the effectiveness of alternative steam generator injection by a portable pump was estimated. We demonstrate the use of this simulation-based dynamic PRA methodology to evaluate the risk induced by a combination of hazards.
Cs concentrations in coastal sediments off FukushimaSuzuki, Shotaro*; Amano, Yosuke*; Enomoto, Masahiro*; Matsumoto, Akira*; Morioka, Yoshiaki*; Sakuma, Kazuyuki; Tsuruta, Tadahiko; Kaeriyama, Hideki*; Miura, Hikaru*; Tsumune, Daisuke*; et al.
Science of the Total Environment, 831, p.154670_1 - 154670_15, 2022/07
Times Cited Count：2 Percentile：29.93(Environmental Sciences)Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 59(3), p.357 - 367, 2022/03
Times Cited Count：5 Percentile：65.59(Nuclear Science & Technology)Dynamic probabilistic risk assessment (PRA), which handles epistemic and aleatory uncertainties by coupling the thermal-hydraulics simulation and probabilistic sampling, enables a more realistic and detailed analysis than 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 station blackout sequence in a boiling water reactor and compared each method. The result indicated that quasi-Monte Carlo sampling method handles the uncertainties most effectively in the assumed scenario.
Kosaka, Wataru; Uchibori, Akihiro; Takata, Takashi; Yanagisawa, Hideki*; Watanabe, Akira*; Jang, S.*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 11 Pages, 2022/03
For the safety assessment of a steam generator (SG) in a sodium-cooled fast reactor, the analysis code LEAP-III can evaluate the water leak rate during the long-term event progress including the tube failure propagation triggered by an occurrence of a small water leak in a failed heat transfer tube in SG. The LEAP-III has the advantage in completing the calculation with low computational cost since it consists of semi-empirical formulae and one-dimensional equations of conservation. However, an evaluation model of temperature distribution by the reacting jet provides wider high temperature region than the experimental data. As a result, LEAP-III shows excessive conservativeness in some case. A Lagrangian particle method code based on engineering approaches has been developed in order to improve this model to get more realistic temperature distribution. In this method, the jet behavior and chemical reaction are simulated using Newton's equation of motion with several engineering approximations instead of solving multi-dimension multiphase thermal hydraulic equations with sodium-water reaction. In this study, interparticle interaction force model was added, and also the chemical reaction and gas-liquid heat transfer evaluation models were improved. We conducted a test analysis, and compared the results by this particle method with the ones by SERAPHIM, that is a mechanistic analysis code for multi-dimensional multiphase flow considering compressibility and sodium-water reaction. Through this test analysis, it confirmed that this particle method has the basic capability to get a realistic temperature distribution with low computational cost, and also to predict tube failure occurrence by coupled with LEAP-III.
Uchibori, Akihiro; Shiina, Yoshimi*; Watanabe, Akira*; Takata, Takashi*
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 12 Pages, 2022/03
An unstructured mesh-based analysis method has been integrated into the sodium-water reaction analysis code, SERAPHIM, in our recent studies. In this study, numerical analysis of an experiment on sodium-water reaction in a tube bundle domain was performed to investigate the effect of the unstructured mesh. The unrealistic behavior appeared in the coarse structured mesh was improved by the unstructured mesh. The numerical result in the case of the unstructured mesh reproduced the peak value of the temperature in the reacting flow.
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.
Kubo, Kotaro; Zheng, X.; Ishikawa, Jun; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 11 Pages, 2020/11
Dynamic probabilistic risk assessment (PRA) enables a more realistic and detailed analysis than classical PRA. However, the trade-off for these improvements is the enormous computational cost associated with performing a large number of thermal-hydraulic (TH) analyses. In this study, based on machine learning (ML), we aim to reduce these costs by skipping the TH analysis. For the ML algorithm, we selected a support vector machine; we built it using a high-fidelity/high-cost detailed model and low-fidelity/low-cost simplified model. As a result, the computational costs could be reduced by approximately 80% without significantly decreasing the accuracy under the assumed conditions.
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.
C thin filmNorimatsu, Wataru*; Matsuda, Keita*; Terasawa, Tomoo; Takata, Nao*; Masumori, Atsushi*; Ito, Keita*; Oda, Koji*; Ito, Takahiro*; Endo, Akira*; Funahashi, Ryoji*; et al.
Nanotechnology, 31(14), p.145711_1 - 145711_7, 2020/04
Times Cited Count：6 Percentile：38.95(Nanoscience & Nanotechnology)We show that boron-doped epitaxial graphene can be successfully grown by thermal decomposition of a boron carbide thin film, which can also be epitaxially grown on a silicon carbide substrate. The interfaces of BC on SiC and graphene on BC had a fixed orientation relation, having a local stable structure with no dangling bonds. The first carbon layer on BC acts as a buffer layer, and the overlaying carbon layers are graphene. Graphene on BC was highly boron doped, and the hole concentration could be controlled over a wide range of 2
10
to 210
cm
. Highly boron-doped graphene exhibited a spin-glass behavior, which suggests the presence of local antiferromagnetic ordering in the spin-frustration system. Thermal decomposition of carbides holds the promise of being a technique to obtain a new class of wafer-scale functional epitaxial graphene for various applications.
Uchibori, Akihiro; Shiina, Yoshimi*; Watanabe, Akira*; Takata, Takashi
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.5315 - 5324, 2019/08
A computational fluid dynamics code SERAPHIM for a sodium-water reaction phenomenon has been developed. An unstructured mesh-based analysis method has been integrated into this code for evaluation in a complex-shaped domain including multiple heat transfer tubes in our recent studies. In this study, a numerical analysis of an experiment on sodium-water reaction in a tube bundle domain was performed to investigate applicability of the unstructured mesh-based SERAPHIM code. The formulated high-temperature region and its temperature level agreed with an experimental result.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Nuclear Technology, 205(1-2), p.119 - 127, 2019/01
Times Cited Count：3 Percentile：31.89(Nuclear Science & Technology)To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.
Li, J.*; Jang, S.*; Yamaguchi, Akira*; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11
The sodium-water reaction model is developed in particle methods. Two chemical reaction model, called surface reaction model and gas-phase reaction model are developed in the particle method. Validation on the case of vapor injection into liquid water is conducted and good consistency of jet velocity evolution along jet trajectory is obtained. Finally, sodium-water chemical reaction in a configuration of multiple tube bundles is simulated. Jet velocity, water vapor fraction and temperature are investigated and reasonable results are observed, which presents promising future of this methodology.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00394_1 - 17-00394_6, 2018/03
For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an underexpanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 54(10), p.1036 - 1045, 2017/10
Times Cited Count：5 Percentile：44.54(Nuclear Science & Technology)To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium. The effect of use of the unstructured mesh was also investigated by the two analyses using structured and unstructured mesh.
Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Watanabe, Akira*
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 12 Pages, 2017/09
To evaluate a sodium-water reaction phenomenon in a steam generator of sodium-cooled fast reactors, a computational fluid dynamics code SERAPHIM, in which a compressible multicomponent multiphase flow with sodium-water chemical reaction is computed, has been developed. The original SERAPHIM code is based on the difference method. In this study, unstructured mesh-based numerical method was developed to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Numerical analysis of an underexpanded jet experiment was performed as part of validation of the unstructured mesh-based numerical method. The calculated pressure profile and location of the Mach disk showed good agreement with the experimental data. Applicability of the numerical method for the actual situation was confirmed through the analysis of water vapor discharging into liquid sodium.
Watanabe, Taro*; Takata, Takashi; Yamaguchi, Akira*
Nuclear Engineering and Design, 313, p.447 - 457, 2017/03
Times Cited Count：2 Percentile：19.65(Nuclear Science & Technology)Countercurrent flow limitation (CCFL) in a heat transfer tube at a steam generator (SG) of pressurized water reactor (PWR) is one of the important issues on the core cooling under a loss of coolant accident(LOCA). In order to improve the prediction accuracy of the CCFL characteristics in numerical simulations using the volume of fluid (VOF) method with less computational cost, a thin liquid film flow in a countercurrent flow is modeled independently and is coupled with the VOF method. Then, we have carried out numerical simulations of a countercurrent flow in a vertical tube so as to investigate the CCFL characteristics and compare them with the previous experimental results. As a result, it has been concluded that the effect of liquid film entrainment by upward gas flux will cause the difference in the experiments.
Kojima, Saori*; Uchibori, Akihiro; Takata, Takashi; Ohno, Shuji; Fukuda, Takeshi*; Yamaguchi, Akira*
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 8 Pages, 2016/11
Analytical evaluation on a self-wastage phenomenon at heat transfer tubes in the steam generator of sodium cooled fast reactors has been performed by using the sodium-water reaction analysis code SERAPHIM. In this study, a fluid-structure thermal coupling model was developed and incorporated in the SERAPHIM code to evaluate heat transfer between the sodium-side reacting flow and the outer surface of the heat transfer tube. The effect of the fluid-structure thermal coupling model on the temperature field was demonstrated through the numerical analyses.
Jang, S.*; Yamaguchi, Akira*; Takata, Takashi
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 11 Pages, 2016/10
The current approach to Level 2 probabilistic risk assessment (PRA) using the conventional event-tree (ET)/fault-tree (FT) methodology requires pre-specifications of event order occurrence and component failure probabilities which may vary significantly in the presence of uncertainties. In the present study, a new methodology is proposed to quantify the level 2 PRA in which the accident progression scenarios are dynamic and interactive with the instantaneous plant state and related phenomena. The accident progression is treated as a continuous Markov process and the transition probabilities are evaluated based on the computation of plant system thermal-hydraulic dynamics. A Monte Carlo method is used to obtain the resultant probability of the radioactive material release scenarios. The methodology is applied to the protected loss of heat sink accident scenario of the level 2 PRA of a generation IV fast reactor.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
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), 11 Pages, 2016/09
For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an under expanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.
