Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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.
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.
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
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-301 and BOP-302R) have been conducted in order to validate the evaluation method of the reactivity feedback equipped in the plant dynamics analysis code named Super-COPD. In this study, 1D-CFD coupled analyses adding the core bowing reactivity model were conducted. Through the analysis, the applicability of the modified reactivity model was confirmed for the BOP-301 test. For the BOP-302R test, consideration of the core restraint system in the core and modeling the control rod driveline expansion reactivity was indicated.
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.
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.
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.
Doda, Norihiro; Ohira, Hiroaki; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1618 - 1625, 2016/04
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method which is required for adoption of natural circulation decay heat removal systems, an analysis of EBR-II (Experimental Breeder Reactor II) shutdown heat removal test using the method was performed. The results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly during natural circulation decay heat removal operations.
Briggs, L.*; Monti, S.*; Hu, W.*; Sui, D.*; Su, G. H.*; Maas, L.*; Vezzoni, B.*; Partha Sarathy, U.*; Del Nevo, A.*; Petruzzi, A.*; et al.
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.3030 - 3043, 2015/08
The International Atomic Energy Agency Coordinated Research Project, "Benchmark Analyses of an EBR-II Shutdown Heat Removal Test" is in the third year of its four-year term. Nineteen participants representing eleven countries have simulated two of the most severe transients performed during the Shutdown Heat Removal Tests program conducted at Argonne's Experimental Breeder Reactor II. Benchmark specifications were created for these two transients, enabling project participants to develop computer models of the core and primary heat transport system, and simulate both transients. In phase 1 of the project, blind simulations were performed and then evaluated against recorded data. During phase 2, participants have refined their models to address areas where the phase 1 simulations did not predict as well as desired the experimental data. This paper describes the progress that has been made to date in phase 2 in improving on the earlier simulations and presents the direction of planned work for the remainder of the project.
Doda, Norihiro; Igawa, Kenichi*; Minami, Masaki*; Iwasaki, Takashi*; Ohira, Hiroaki
no journal, ,
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method, which is required for adoption of natural circulation decay heat removal systems, EBR-II (Experimental Breeder Reactor II) shutdown heat removal test was simulated. The simulation results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly in natural circulation decay heat removal.
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.
Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Uwaba, Tomoyuki; Tanaka, Masaaki; Nemoto, Toshiyuki*
no journal, ,
A sodium-cooled fast reactor has an inherent safety feature of feedback reactivity. Core deformation reactivity decreases fission power automatically in case of increase of the reactor power due to the negative reactivity according to raise of the core temperature. To improve the evaluation accuracy of the core deformation reactivity, deflection of the core support plate which varies the inclination of fuel assemblies and the pitches among them at the center height of the core and has impact on the reactivity was investigated quantitatively in the high flowrate and low flowrate conditions separately by structural mechanics analyses.
Doda, Norihiro; Kato, Shinya; Yoshimura, Kazuo; Uwaba, Tomoyuki; Yokoyama, Kenji; Tanaka, Masaaki
no journal, ,
JAEA has developed an evaluation method coupling neutronics, core thermal-hydraulics, and core structural mechanics to evaluate more realistically core deformation reactivity due to reactor power variation during anticipated operational occurrence or accidents in sodium-cooled fast reactors. The analysis results of the EBR-II SHRT-45R test simulating a loss of flow event using the evaluation method showed that the core deformation reactivity was negative from the start. When the core temperature rose, fuel assemblies bowed outward in the core due to the difference in thermal expansion between the facing wrapper walls of the assembly, causing negative reactivity in the fuel region and the surrounding reflector region due to the displacement of fuel and structural materials. As a result, the availability of the core deformation reactivity evaluation method was confirmed.
