Sodium-cooled fast reactors have intrinsic safety features decreasing reactor power during the increase of the core inlet temperature by the feedback reactivity due to the core deformation. 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, to enable the plant dynamics analyses taking account of the thermal stratification in the cold pool, the 1D-CFD coupling method in which CFD was applied to the cold pool was applied to the ULOHS (Unprotected Loss Of Heat Sink) test performed in the experimental fast reactor EBR-II in the U. S. and the evaluation of the core inlet temperature could be improved. By using this 1D-CFD coupling method, sensitivity analyses concerning the core bowing reactivity were carried out with the aim of improving the evaluations of the core deformation reactivity. Through the numerical analyses, the applicability of the core bowing reactivity model to the test could be indicated.
To achieve an innovative core design, an optimization process for a core design has been developed as a part of the design optimization support tool named ARKADIA-Design. The core design optimization process integrated the core design analysis of neutronics, thermal-hydraulics, and fuel integrity and plant dynamics analysis with the Bayesian optimization (BO) algorithm is being developed. The BO can reduce the total number of iterative calculations and enhance the efficiency of the optimization. To establish a basic framework of the optimization process, a representative problem following an actual core design procedure was defined. Here, core design parameters are optimized to show a high core performance with inherent safety by preventing core damage in an unprotected loss of flow event of sodium-cooled fast reactors. In this study, to confirm an applicability of the optimization process with the BO algorithm for the representative problem, a single-objective optimization problem was solved by performing the integrated analysis only between neutronics and plant dynamics as a first trial. In addition, an effectiveness of the optimization process was discussed by comparing with an ordinary core design process. An optimal solution in the representative problem was acquired by performing the integrated analysis with the constraint BO.
Abstract In order to enhance the safety of sodium-cooled fast reactors, a direct reactor auxiliary cooling system (DRACS) under natural circulation conditions with a dipped-type direct heat exchanger (D-DHX) in an upper plenum of the reactor vessel has been investigated. During the DRACS operation, the complicated thermal-hydraulic phenomena that cold coolant from D-DHX flowed into the fuel subassemblies and narrow gaps between them, which is well-known as inter-wrapper flow (IWF) was observed. Therefore, a multi-dimensional thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics (CFD) code (RV-CFD model) has been developed to evaluate the core cooling performance under natural circulation conditions. For the design study, the RV-CFD model is demanded to simulate with reasonable calculation costs while maintaining accuracy. In this paper, the application of the subchannel analysis method by CFD code for fuel subassemblies (subchannel CFD model) to the RV-CFD model was attempted. In the subchannel CFD model, the porous media approach was used to consider local geometry in the fuel subassembly, and the effective heat conductivity coefficients in diffusion term of the energy equation were set to fit the actual radial thermal diffusion between subchannels. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1 and the calculated sodium temperature in the core had good agreement with the experimental result. It was confirmed that the RV-CFD model with subchannel CFD model was applicable to the core thermal-hydraulic analysis during the DRACS with the D-DHX operation under natural circulation conditions.
To enhance a safety of sodium-cooled fast reactors (SFRs), decay heat removal systems under natural circulation (NC) with a dipped-type direct heat exchanger (D-DHX) installed in an upper plenum of a reactor vessel (RV) have been investigated. During the D-DHX operation, the flow of the coolant at low temperature from the D-DHX into assemblies and an interwrapper gap (IWG) between them, and the radial heat transfer through a wrapper tube and the IWG among assemblies occur. Such phenomena in the core can remove the decay heat without external electric power supply. In terms of the design study, modeling of an RV using a CFD code (RV-CFD) with a coarse mesh arrangement has an advantage in a reduced computational cost. In this study, focused on the modeling of the IWG, to achieve a lower computational cost while maintaining the prediction accuracy, an influence of combination of the mesh number in the IWG and the pressure loss correlation on the core temperature distribution was investigated through the numerical analysis of a sodium experimental apparatus named PLANDTL-1. The result shows the coarse mesh with correlation reduced the IWF, or a circulation flow with the upward and downward flow in the IWG, and shifted the temperature distribution in the core to the high-temperature side.
A natural circulation evaluation methodology has been developed to insure safety of a sodium cooled fast reactor (SFR) of 1500 MWe adopting a natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can be applied to safety evaluation for SFR licensing taking into account the temperature flattening effect due to buoyancy force in the core, and a three-dimensional fluid flow analysis which can evaluate thermal-hydraulics for local convection and thermal stratification in the primary system and DHRSs. The one-dimensional safety analysis method and the three-dimensional fluid flow analysis method have been validated using the test results of a water test apparatus and a sodium test loop for some typical transient events selected from the design basis events of the SFR. Finally, it has been confirmed that a good agreement between the test results and analysis results has been obtained, and reliability of each method has been demonstrated.
A plant dynamic analysis code named Super-COPD have been developing for safety and design studies of a sodiumcooled fast reactor. Validation of the plant analysis model including neutronics calculation of a one-point reactor kinetics model against experimental data necessitates the further work on the beyond design basis accident to improve the prediction accuracy. Therefore, Japan Atomic Energy Agency participated in International Atomic Energy Agency (IAEA) benchmark for Loss of Flow without Scram (LOFWOS) test No.13 performed at the Fast Flux Test Facility (FFTF) as one of the passive safety demonstration tests. The transient analysis considering with major reactivity feedback mechanisms was carried out with the Super-COPD and the results at the first blind analysis phase of the benchmark project were assessed against the measured data. It was observed that the whole plant dynamics analysis by the SuperCOPD could follow the measured data and the analysis model had the prospect of applicability to the LOFWOS event in the beyond design basis accident. However, discrepancies were still observed in the outlet sodium temperature profile of two fast-response proximity instrumented open test assemblies (PIOTAs) in the natural circulation conditions. As a future work, the core model will be refined to improve the prediction accuracy of radial heat transfer rate.
A natural circulation evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analysis adopting the turbulence model. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.
