In a probable scenario for core disruptive accidents of Sodium-cooled Fast Reactors (SFRs), it is foreseen that molten core material would be discharged into lower sodium plenums through control rod guide tubes. Such material relocation might lead to a considerable thermal load on lower structures of the reactor vessels, while it has been suggested that in SFRs, as soon as the molten core material is discharged into coolant, it might be fragmented into smaller particles by fuel-coolant interactions and thus efficiently cooled in the reactor vessels. Hence, understanding of the fragmentation is crucial for achieving in-vessel retention of molten core material in SFRs. In this paper, based on the experimental results of a series of fragmentation tests, where around 10 kg of molten alumina (Al 2 O 3 ) was discharged into a sodium pool (depth: 1.3 m, diameter: 0.4 m, temperature: 673 K) through a duct (inner diameter: 40mm to 63 mm) by using an experimental facility at National Nuclear Center of the Republic of Kazakhstan, dominant mechanisms for the fragmentation are discussed. In the present tests, mass median diameters of solidified Al 2 O 3 particles were around 0.3 mm, which were comparable to the values predicted using conventional hydrodynamic-instability theories. However, even though the conventional theories predict that particle size becomes smaller with the increase of Weber number, such tendency was not observed in the present tests. Taking into account that in the present tests, the distances for fragmentation of molten Al 2 O 3 were evaluated to be approximately 60 % to 70 % below the values predicted using an existing representative correlation which regards hydrodynamic instabilities as a dominant fragmentation mechanism, the observed independence on Weber number confirms a mechanism that before hydrodynamic instabilities sufficiently grow to induce fragmentation, thermal phenomena such as local coolant vaporization and resultant vapor expansion significantly accelerate fragmentation in SFRs.
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.
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.
In the design of a Japanese sodium-cooled fast reactor (JSFR), a design measure (fuel subassembly with an inner duct structure; FAIDUS) is considered to prevent severe recriticality events even in the case of core disruptive accidents by molten fuel ejection out of the core region through the duct equipped within the fuel subassembly. Conˆrming the principle eŠectiveness of such a design measure is important. In this study, the systematic heat transfer behavior in the ID1 test, which was conducted in the impulse graphite reactor (IGR) in Republic of Kazakhstan, was evaluated by applying a heat conduction code TAC2D and a reactor safety analysis code SIMMER III focusing on the clariˆcation of heat transfer from a hightemperature mixture of molten fuel and steel to the duct. As a result, the duct failure caused by high heat ‰ux from the mixture was identiˆed as one of an important mechanisms of early duct failure in FAIDUS. It was also suggested from this study that the high heat ‰ux from the mixture is caused by the direct contact of molten steel in the absence of fuel crust on the duct wall. Based on theseˆndings, it is judged that the mechanism of early duct failure with high heat ‰ux obtained in the ID1 test satisˆes the required condition for FAIDUS, i.e., the inner duct of FAIDUS should fail at an early phase of core disruptive accident in advance of wrapper tube failure so that the produced molten fuel can escape from the core region, which supports the feasibility of the FAIDUS concept.
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