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Ohshima, Hiroyuki; Asayama, Tai; Furukawa, Tomohiro; Tanaka, Masaaki; Uchibori, Akihiro; Takata, Takashi; Seki, Akiyuki; Enuma, Yasuhiro
Journal of Nuclear Engineering and Radiation Science, 9(2), p.025001_1 - 025001_12, 2023/04
This paper describes the outline and development plan for ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant, including optimization of safety equipment. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-Design and ARKADIA-Safety will be constructed individually, with the first target of sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.
Uchibori, Akihiro; Sonehara, Masateru; Aoyagi, Mitsuhiro; Takata, Takashi*; Ohshima, Hiroyuki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
A new computational code, SPECTRA, has been developed for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors. The in-vessel thermal hydraulics module includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.
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, 11 Pages, 2022/04
Times Cited Count：1 Percentile：0.02(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, 59(3), p.357 - 367, 2022/03
Times Cited Count：2 Percentile：44.81(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.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 13 Pages, 2022/03
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.
Kosaka, Wataru; Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Jang, S.*
Nihon Kikai Gakkai Rombunshu (Internet), 88(905), p.21-00310_1 - 21-00310_9, 2022/01
If a pressurized water/water-vapor leaks from a heat transfer tube in a steam generator (SG) in a sodium-cooled fast reactor (SFR), sodium-water reaction forms high-velocity, high-temperature, and corrosive jet. It would damage the other tubes and might propagate the tube failure in the SG. Thus, it is important to evaluate the effect of the tube failure propagation for safety assessment of SFR. The computational code LEAP-III can evaluate water leak rate during the tube failure propagation with short calculation time, since it consists of empirical formulae and one-dimensional equations of conservation. One of the empirical models, temperature distribution evaluation model, evaluates the temperature distribution in SG as circular arc isolines determined by experiments and preliminary analyses instead of complicated real distribution. In order to improve this model to get more realistic temperature distribution, we have developed the Lagrangian particle method based on engineering approaches. In this study, we have focused on evaluating gas flow in a tube bundle system, and constructed new models for the gas-particles behavior around a tube to evaluate void fraction distribution near the tube. Through the test analysis simulating one target tube system, we confirmed the capability of the models and next topic to improve the models.
Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 15 Pages, 2022/00
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.
Zhan, Y.*; Sun, G.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi
Experimental Thermal and Fluid Science, 126, p.110402_1 - 110402_8, 2021/08
Times Cited Count：1 Percentile：35.04(Thermodynamics)Takata, Takashi; Aoyagi, Mitsuhiro; Sonehara, Masateru
IAEA-TECDOC-1972, p.224 - 234, 2021/08
Sodium fire is one of the key issues for plant safety of sodium-cooled fast reactor (SFR) regardless of its size. In general, a concrete structure, which includes free and bonging water inside, is used in a reactor building. Accordingly, water vapor will release from the concrete during sodium fire incident due to temperature increase resulting in a hydrogengeneration even in a dry air condition. The probability of hydrogen generation will increase in accordance with a decrease of a dimension of compartment that corresponds to a small and medium sized or modular reactor (SMR). A numerical investigation of a small leakage sodium pool fire has been carried out by changing a dimension of compartment. Furthermore, numerical challenges to enhance a prediction accuracy of hydrogen generation during sodium fire has also been discussed in the paper.
Aoyagi, Mitsuhiro; Takata, Takashi; Uno, Masayoshi*
Nuclear Engineering and Design, 380, p.111258_1 - 111258_11, 2021/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Sugimoto, Taro*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Kurihara, Akikazu; Takata, Takashi; Ohshima, Hiroyuki
Nuclear Engineering and Design, 380, p.111306_1 - 111306_11, 2021/08
Times Cited Count：2 Percentile：68.98(Nuclear Science & Technology)Liquid droplet entrainment by a high-speed gas jet is a key phenomenon for evaluation of sodium-water reaction. In this study, a visualization experiment for liquid droplet entrainment by an air jet in a water pool by using frame-straddling method was carried for development of an entrainment model in a sodium-water reaction analysis code. This experiment successfully provided clear images that captured generation and movement of droplets. Droplet diameter and moving speed were obtained at different locations and gas jet velocities from image processing. The measured data contributes phenomena elucidation and model development.
Kosaka, Wataru; Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Jang, S.*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 6 Pages, 2021/08
For safety assessment or design of a steam generator (SG) of a sodium-cooled fast reactor, it is important to evaluate the effects of a multiphase flow involving sodium-water reaction. If pressurized water/water-vapor leaks from a tube, it forms a corrosive, high-temperature, and high-velocity jet, and may cause failure of the adjacent tubes. The occurrence of tube failure on many tubes will lead to failure of the boundary between the primary and secondary cooling loops. The numerical analysis code, LEAP-III, has been developed to evaluate water leak rate considering the effects of the above-mentioned phenomena with short computational time. In some cases, however, the current LEAP-III provides excessive conservativeness due to its temperature distribution evaluation model. In order to reduce this excess, we have developed a new Lagrange particle method with several engineering approaches. We also performed test analyses which simulate time development of the vapor jet with chemical reaction in a SG. The results of the developed method were compared with ones of the multi-dimensional multiphase thermal hydraulic analysis code, SERAPHIM which considers compressibility and chemical reaction. Through the test analyses, the basic capability of the developed method was confirmed.
Aoyagi, Mitsuhiro; Louie, D. L. Y.*; Uchibori, Akihiro; Takata, Takashi; Luxat, D.*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 10 Pages, 2021/08
Saito, Masafumi*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Kurihara, Akikazu; Takata, Takashi*; Ohshima, Hiroyuki
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 7 Pages, 2021/08
In order to provide the data for validation and improvement of the sodium-water reaction analysis code, a visualization experiment on liquid droplet entrainment in a high-pressure air jet submerged in a water pool was conducted. Diameter and velocity of entrained liquid droplets were successfully measured. The effect of a nozzle shape was elucidated.
Kosaka, Wataru; Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Jang, S.*
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07
If a pressurized water/water-vapor leaks from a heat transfer tube in a steam generator (SG) in a sodium-cooled fast reactor (SFR), sodium-water reaction forms high-velocity, high-temperature, and corrosive jet. It would damage the other tubes and might propagate the tube failure in the SG. Thus, it is important to evaluate the effect of the tube failure propagation for safety assessment of SFR. The computational code LEAP-III can evaluate water leak rate during the tube failure propagation with short calculation time, since it consists of empirical formulae and one-dimensional equations of conservation. One of the empirical models, temperature distribution evaluation model, evaluates the temperature distribution in SG as circular arc isolines determined by experiments and preliminary analyses instead of complicated real distribution. In order to improve this model to get more realistic temperature distribution, we have developed the Lagrangian particle method based on engineering approaches. In this study, we have focused on evaluating gas flow in a tube bundle system, and constructed new models for the gas-particles behavior around a tube to evaluate void fraction distribution near the tube. Through the test analysis simulating one target tube system, we confirmed the capability of the models and next topic to improve the models.
Li, C.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07
The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.
Sonehara, Masateru; Uchibori, Akihiro; Aoyagi, Mitsuhiro; Kawada, Kenichi; Takata, Takashi; Ohshima, Hiroyuki
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2021/07
In sodium-cooled fast reactors (SFRs), it has been pointed out that molten fuel may be discharged from the core during a severe accident (SA) accompanied by core damage, and may solidify into debri particles with diameters ranging from several millimeters to several hundred micrometers due to interaction with the sodium coolant and accumulate at the bottom of the reactor vessel. Therefore, it is necessary to understand the behavior of such debri particles appropriately to evaluate the SA event progression. To meet these requirements, a molten fuel behavior analysis code using dissipative particle dynamics (DPD), a kind of particle method, has been developed as a part of the SPECTRA code, tool for consistent analysis of in-vessel and ex-vessel events in sodium fast reactor accidents. In this study, it was found that the new analyses code can reproduce sedimentation behavior of particles by adding a new stress term in the shear direction.
Miura, Daisuke*; Kumada, Takayuki; Sekine, Yurina; Motokawa, Ryuhei; Nakagawa, Hiroshi; Oba, Yojiro; Ohara, Takashi; Takata, Shinichi; Hiroi, Kosuke; Morikawa, Toshiaki*; et al.
Journal of Applied Crystallography, 54(2), p.454 - 460, 2021/04
Times Cited Count：0 Percentile：0.01(Chemistry, Multidisciplinary)We developed a spin-contrast-variation neutron powder diffractometry technique that extracts the structure factor of hydrogen atoms, namely, the contribution of hydrogen atoms to a crystal structure factor. Crystals of L-glutamic acid were dispersed in a dpolystyrene matrix containing 4-methacryloyloxy-2,2,6,6,-tetramethyl-1-piperidinyloxy (TEMPO methacrylate) to polarize their proton spins dynamically. The intensities of the diffraction peaks of the sample changed according to the proton polarization, and the structure factor of the hydrogen atoms was extracted from the proton-polarization dependent intensities. This technique is expected to enable analyses of the structures of hydrogen-containing materials that are difficult to determine with conventional powder diffractometry.
