In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progressing beyond the fuel subassembly scale and that the upward discharge option will be effective in eliminating the energetic potential.

In the safety analysis of sodium-cooled fast reactors, numerical simulations of various thermal-hydraulic phenomena with multicomponent and multiphase flows in core disruptive accidents (CDAs) are regarded as particularly difficult. In the material relocation phase of CDAs, core debris settle down on a core support structure and/or an in-vessel retention device and form a debris bed. The bed’s shape is crucial for the subsequent relocation of the molten core and heat removal capability as well as re-criticality. In this study, a hybrid numerical simulation method, coupling the multi-fluid model of the three-dimensional fast reactor safety analysis code SIMMER-IV with the discrete element method (DEM), was applied to analyze the sedimentation and bed formation behaviors of core debris. Three-dimensional simulations were performed and compared with results obtained in a series of particle sedimentation experiments. The present simulation predicts the sedimentation behavior of mixed particles with different properties as well as homogeneous particles. The simulation results on bed shapes and particle distribution in the bed agree well with experimental measurements. They demonstrate the practicality of the present hybrid method to solid particle sedimentation and bed formation behaviors of mixed as well as homogeneous particles.

The elimination of severe power excursion during core disruptive accidents is a key issue for the enhanced safety of sodium-cooled fast reactors. In order to prevent the formation of a large-scale molten fuel pool within a reactor core, which is one of the factors leading to the severe power excursion, the Japan Atomic Energy Agency (JAEA) is considering the introduction of fuel assembly with inner duct structure (FAIDUS). In the current reference design for FAIDUS, the top end of the inner duct is open, whereas the bottom end is closed, and therefore it is expected that the molten fuel will be discharged from the reactor core towards the upper sodium plenum through the inner duct. The objective of the present study is to clarify the fundamental mechanism for upward fuel discharge through the inner duct structure, and thereby to confirm the effectiveness of FAIDUS. The possibility of upward discharge of a high-density melt driven by coolant vapor has been confirmed by the JAEA's experiment, in which molten Wood's metal simulating the molten fuel was injected into a coolant channel (equivalent inner diameter: 30 mm, total height: 2 m, fluid content: water) simulating the inner duct structure. In this paper, the mechanism of upward discharge of a high-density melt driven by coolant vapor pressure and/or flow in this experiment is discussed in terms of the application to reactor conditions. Through this discussion, the following mechanisms were clarified. (1) Coolant vapor pressure is built up within the coolant channel after the melt injection. The magnitude of the pressure buildup becomes larger with the increase of melt-enthalpy-injection rate, which is defined by the product of melt-mass-injection rate into the coolant channel and melt specific enthalpy. (2) Following the pressure buildup, the melt is discharged upward, being driven by the coolant vapor flow directing towards the top opening end of the coolant channel. The upward discharge mass rate becomes higher with the increase of the magnitude of the pressure buildup and, therefore, the melt-enthalpy-injection rate. The experimental knowledge obtained from the JAEA's experiment suggests that the coolant pressure buildup could act as one of the driving forces for the upward discharge of a high-density melt through the inner duct structure in FAIDUS under reactor conditions with higher melt-enthalpy-injection rate than the current simulant experimental condition.

In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropriate design measures.

Analyses of severe accidents for sodium-cooled fast reactors have shown that by assuming pessimistic conditions the accident might proceed into a transition phase where a large whole-core-scale pool containing sufficient fuel to exceed prompt criticality by fuel compaction might be formed. Local fuel-coolant interaction (FCI) in the pool is regarded as one of the probable initiators that could lead to such compactive fluid motions. To enhance the evaluation of severe accidents, a three-step research plan has been determined. In Phase 1, a series of simulated experiments, which covers a variety of conditions including much difference in water volume, melt temperature, water subcooling and water release site, was conducted by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy, while in Phase 2, the interaction characteristics (esp. the pressure buildup) was investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It was recognized that the SIMMER-III code, with its existing heat-transfer models, can reasonably reproduce the experimental evidence observed, thereby greatly stimulating us to initiate the Phase 3, in which further numerical analyses using reactor materials are planned. In this work, based on the latest calculations, it is confirmable that similar to the observations in Phases 1 and 2, for a given melt and sodium temperature within the non-film boiling condition, as the volume of sodium entrapped within the pool increases, a limited pressure-buildup is achievable. In addition, the performed analyses also suggest that despite of a comparatively larger temperature range of molten-fuel and sodium possibly varied during reactor accident progression, the isolation effect of vapor bubbles generated at the melt-sodium interface seems to be the unique dominant mechanism that leads to such limited pressurization. Knowledge and fundamental data from this work might be utilized for future empirical-approach studies (e.g. a lookup table for estimating the critical coolant volume required for achieving the saturated pressurization at varied melt and coolant temperatures).