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Horiguchi, Naoki; Yoshida, Hiroyuki; Kaneko, Akiko*; Abe, Yutaka*
Nihon Kikai Gakkai Kanto Shibu Dai-29-Ki Sokai, Koenkai Koen Rombunshu (Internet), 5 Pages, 2023/10
To elucidate the behavior of molten fuels as a liquid jet in a shallow pool, which is assumed in a core meltdown accident of an LWR, and develop the evaluation method, we investigated the behavior of the vortical liquid film of the simulated wall-impinging liquid jet using 3-dimensional interface shape data obtained by the experiment in a liquid-liquid system.
4 simulated fuel bundle for validation of thermal-hydraulics simulation codesNagatake, Taku; Shibata, Mitsuhiko; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
JAEA is developing a neutronics/thermal-hydraulics coupling simulation code for light-water reactors. Thermal-hydraulic simulation codes applied to the platform are expected to evaluate void fraction distributions in fuel assemblies under operational conditions, which is necessary for neutron transport simulation, and need to be validated using void fraction distribution data in a rod bundle under high-temperature and high-pressure conditions. To obtain the data for code validation, we have been measuring the instantaneous void fraction distribution in a 44 simulated fuel assembly by a wire mesh sensor. In this paper, we report the results of the experiments with pressure and flow rate as parameters at a maximum pressure of 2.6 MPa.
Yamano, Hidemasa; Kurisaka, Kenichi; Takano, Kazuya; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Shirakura, Shota*
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. The project overview is presented in this report.
Ishida, Shinya; Tagami, Hirotaka; Tobita, Yoshiharu; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
no abstracts in English
Ishii, Katsunori; Morita, Keisuke; Noguchi, Hiroki; Aoki, Takeshi; Mizuta, Naoki; Hasegawa, Takeshi; Nagatsuka, Kentaro; Nomoto, Yasunobu; Shimizu, Atsushi; Iigaki, Kazuhiko; et al.
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2023/09
Okajima, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2023/09
no abstracts in English
Ezure, Toshiki; Akimoto, Yuta; Matsushita, Kentaro; Tanaka, Masaaki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
In hot plenums of sodium-cooled fast reactors, restriction of cover gas entrainment caused by vortex dimples on the free surface is an important thermal-hydraulic issue. For this reason, the authors have developed an evaluation method of gas entrainment with an evaluation tool named, StreamViewer. In this study, evaluation using StreamViewer was applied to a water experiment having a simplified hot pool geometry aiming at the validation of the evaluation method toward the application to the evaluation of a pool-type sodium cooled fast reactor. In StreamViewer, the three-dimensional distribution of pressure decrease along the vortex center line was calculated from the velocity distribution obtained by CFD analyses, and the free surface dimple depth was obtained from the hydraulic balance with the pressure distribution and the cover gas pressure. As the results, it was confirmed that the onset of gas entrainment could be predicted appropriately based on the above-mentioned calculation method.
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.
Yoshimura, Kazuo; Doda, Norihiro; Nakamine, Yoshiaki*; Fujisaki, Tatsuya*; Igawa, Kenichi*; Iida, Masaki*; Tanaka, Masaaki
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
In Japan Atomic Energy Agency, a virtual plant model of the sodium-cooled fast reactor plant composed in a computer is being developed to reduce the development cost, by replacing the experiments to the numerical simulations with coupled analyses of the physical phenomena accounting for the interaction between components under various plant conditions. Through the numerical analysis of the ULOHS test in the U.S. experimental fast reactor named EBR-II, applicability of the virtual plant model was confirmed in comparison with the measured data including the core inlet temperature and the reactor power.
Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Nishi, Tsuyoshi*; Morita, Koji*; Nakamura, Kinya*; Pellegrini, M.*
Nihon Kikai Gakkai 2023-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2023/09
A research project has been conducting thermophysical property measurement of a eutectic melt, eutectic melting reaction and relocation experiments, eutectic reaction mechanism investigation, and physical model development on the eutectic melting reaction for reactor application analysis in order to simulate the eutectic melting reaction and relocation behavior of boron carbide as a control rod material and stainless steel during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan. This paper describes the project overview and progress until JFY2022.
Noi, Hiromi*; Watanabe, Sota*; Kubo, Koji*; Okajima, Satoshi; Ando, Masanori
Nihon Kikai Gakkai M&M 2023 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.CL0712_1 - CL0712_5, 2023/09
no abstracts in English
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.
Yamano, Hidemasa; Morita, Koji*
Nihon Kikai Gakkai 2022-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2022/09
It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B4C) as a control rod material and stainless steel (SS) during a core disruptive accident (CDA) in an advanced large-scale sodium-cooled fast reactor (SFR) designed in Japan. A physical model simulating the eutectic reaction and relocation of the eutectic melt was developed to incorporate into the fast reactor severe accident analysis code SIMMER-IV for the CDA numerical analysis of SFRs. This study applied the SIMMER-IV code with the newly developed model to the CDA analysis of the SFR. This analysis indicated that the SIMMER-IV code using the eutectic reaction model has successfully simulated the eutectic reaction and the upward motion of the eutectic melt in the molten core pool as well as the reactivity transient behavior caused by the molten core material relocation.
Ono, Ayako; Yamashita, Susumu; Sakashita, Hiroto*; Suzuki, Takayuki*; Yoshida, Hiroyuki
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
JAEA is implementing a simulation of a two-phase flow in the reactor core by TPFIT and JUPITER which are developed by JAEA based on the surface tracking method. However, it is impossible to simulate a boiling on the heating surface in the large-scale domain by this type of simulation method since the simulation of boiling based on the surface tracking method needs the fine mesh which sufficiently resolves the initiation of boiling. Therefore, JAEA started to develop the simplified boiling model applied for the two-phase flow in the fuel assemblies. In this study, the simulation results of the convection boiling on a vertical heating surface and the comparison between the simulation results and experimental results are shown.
Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Akayev, A. S.*; Mikisha, A. V.*; Baklanov, V. V.*; Vurim, A. D.*
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
The cooling of the residual core materials after the fuel discharge from the SFR core in the core disruptive accident can significantly affect the distribution fraction of the core materials which is an important factor for the in-vessel retention (IVR). The cooling of the residual core materials is called "in-place cooling". For the evaluation of the in-place cooling, behavior in a SFR core was simulated by SIMMER-III, and method of phenomena identification and ranking table (PIRT) was applied based on the analysis result. Experiment which focuses on the thermal-hydraulic phenomena which were extracted by the PIRT was conducted in the framework of EAGLE-3 project. Continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs was observed in the experiment, and analysis by the SIMMER-III was conducted. By investigation of the analysis result, difference between the experiment and analysis results was revealed to be due to remaining and occupation of non-condensable gas above the sodium level which would be unrealistic in the experiment. Gas mixture model between non-condensable gas and sodium vapor was developed to solve this problem, and coincidence between experiment and analysis results was largely improved by this new model.
Emura, Yuki; Kamiyama, Kenji; Yamano, Hidemasa
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
no abstracts in English
Kikuchi, Norihiro; Mori, Takero; Okajima, Satoshi; Tanaka, Masaaki; Miyazaki, Masashi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
In JAEA, the design optimization method for plant structure has been developed on the process to output optimal solution of the thickness of reactor vessel wall against thermal transient and seismic loads in a SFR as a representative problem. Resistance characteristic of the wall on the load derived from thermal transient is one of the most important factors for safety estimation on the structural integrity. Failure probability of components against thermal transient was set to one of variables in the objective function for a common scale to compare with other variables in different failure mechanisms. In the iterative process to achieve the optimal solution, a number of evaluations to measure the influence on the load derived from thermal transient was necessarily conducted. More reduction of required time for evaluations is desired. To perform the iterative evaluation process efficiently, the automatization of parametric analyses was implemented in the optimization process.
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
Identifying accident scenarios that could lead to severe accidents and evaluating their frequency of occurrence are essential issues. This study aims to establish the methodology of the dynamic Probabilistic Risk Assessment (PRA) for sodium-cooled fast reactors that can consider the time dependency and the interdependence of each event. Specifically, the Continuous Markov chain Monte Carlo (CMMC) method is newly applied to the SPECTRA code, which analyzes the severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. Currently, a fault-tree model of air coolers of decay heat removal system is implemented as the CMMC method, and a series of preliminary analysis of the plant's transient characteristics under the scenario of volcanic ashfall has been conducted.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Matsushita, Kentaro; Ezure, Toshiki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Tanaka, Masaaki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2021-Nendo Koen Rombunshu (Internet), 4 Pages, 2021/08
For evaluation of gas entrainment phenomenon at free surface in reactor vessel of sodium-cooled fast reactor, the gas entrainment evaluation tool named "Stream Viewer" has been developed. In Stream Viewer, depth of surface vortex dimple is predicted by calculating pressure decrease at the vortex center using velocity distribution around the vortex and Burgers vortex model. In this report, a method to identify continuous vortex center lines from a velocity distribution is newly developed. It becomes possible to evaluate three-dimensional distribution of pressure decrease along vortex center line. Then, the method is validated by applying Stream Viewer to an open channel experiment. As the result, it was confirmed that vortex center lines were successfully identified by the improved Stream Viewer. Moreover, it was also shown that the evaluation accuracy of gas entrainment was expected to be improved by considering distribution of pressure decrease along vortex center line.
