Conference Agenda

In the frame of the French ANR Post-Fukushima ICE program (“Interaction between Corium and Water”), a series of Fuel Coolant Interaction (FCI)/Ex- vessel Steam Explosion (EVSE) integral tests have been performed at the KROTOS facility of the Severe Accident platform PLINIUS, located at CEA-Cadarache. Complementary, thermodynamic and thermophysical properties of prototypic corium chosen for integral tests KROTOS have been measured on VITI facility (CEA-Cadarache) and ATTILHA facility (CEA-Saclay). Post-test analyses on corium steam exploded debris have been performed through high-resolution X-ray diffraction measurements done on the MARS beamline at the SOLEIL synchrotron radiation source. The experimental research of ICE program was focused on fragmentation, dispersion of corium jets and formation of debris beds mechanisms, steam explosion energetics, corium oxidation and solidification mechanisms. The up-grade of KROTOS facility and new configurations for VITI and ATTHILA facilities to answer to the ICE program scientific objectives will be presented. Three integral KROTOS tests will be described and the knowledge gained for FCI/SE modelling will be discussed. A special focus will be done in the assessment of thermodynamic and thermophysical corium properties measurements and modelling. The very new results obtained concerning the corium final solid state and the cationic composition fluctuation that occurs in the U_1-xZr_xO_2-y solid solution will be presented.

Advanced Technology Fuels (ATF) cladding materials have become a key research focus worldwide, particularly at KIT, due to their promising potential to enhance reactor safety under accident conditions. FeCrAl and Cr-coated Zirconium alloys are designed to reduce hydrogen generation at least at the beginning of severe accidents (SAs). Their application in Small Modular Reactors (SMRs) is particularly relevant, as SMRs are emerging as a safer alternative to traditional Nuclear Power Plants (NPPs) due to their reduced core inventory and advanced safety systems. Within the framework of the EU SASPAM-SA project, this study focuses on the analysis of hypothetical SA scenarios involving a generic integral Pressurized Water Reactor (iPWR) with natural circulation, using the integral code ASTEC v3.1.2, ASNR all rights reserved, 2024. This code models thermohydraulic and physicochemical phenomena, allowing a detailed assessment of the accident progression from the initial event to the potential release of the radioactive material to the environment. This study specifically examines the performance of Zircaloy-4 and ATF cladding materials, focusing on their influence on the core degradation and hydrogen generation. The results show distinct hydrogen release kinetics for ATF materials compared to Zircaloy, emphasizing the impact of cladding properties on hydrogen production and safety margins during SAs.

The optimization of post-accident management in case of a severe accident (SA) is complex, particularly in what concerns handling of nuclear materials. After a SA, most of nuclear materials remain within the nuclear power plant (NPP) units in a solidified state, usually referred to as corium. The dismantling phase entails cutting such corium chunks into manageable pieces without causing an unnecessary radioactive remobilization to the gas phase that might result in further source term to the environment. This is currently the stage to be faced shortly in the Fukushima Daiichi site.

The OECD/FACE (Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation) project is dedicating significant resources to finding the best process for dismantling the site. Achieving this goal requires exploring different techniques and protection measures. This work is an exploratory analysis on how effectively an overlying water layer could absorb particulate material generated during laser cutting, in preparation for retrieving fuel debris from the affected units. Using the SPARC-Jet code, an in-house extension of SPARC-90 (Suppression Pool Aerosol Removal Code), the influence of uncertain boundary conditions on water retention efficiency has been studied. The focus was on factors such as carrier gas flow rate, particle size and concentration, pool temperature, and water depth. Preliminary results suggest that injection spot diameter and injected gas mass flow rate lead to higher Decontamination Factor (DF) values. However, from the cleaning efficiency standpoint, variables such as water temperature or depth should not be a concern, as their effect is very minor.

This work is being carried out in the context of severe accidents mitigation in sodium-cooled fast reactors (SFRs). The corium (set of molten core materials) formed may be transferred through discharge tubes to the lower part of the reactor vessel, towards a core-catcher. However, this corium could reach the core-catcher in the form of a hot jet (3000K), which could lead to local ablation of the core-catcher. This risk must therefore be taken into account to ensure that the core-catcher retains its integrity during corium relocation phase.

In the present work, the effect of the core-catcher geometry on its ablation process by a hot jet is investigated experimentally. Experiments were conducted on HAnSoLO setup, with simulating materials (transparent ice /jet of water). The experimental conditions were determined to be as representative as possible of those of a nuclear reactor. The geometric features of the core-catcher which are studied are its inclination and roughness. It has been observed that these two parameters significantly influence the ablation phenomenon, and in some cases can increase the ablation rate.

Sodium Cooled Fast Reactor (SFR) is one of the most promising Gen-IV concept for earliest deployment, owing to vast operating experience worldwide. Currently operating SFRs adapted partial core meltdown as a design basis for Core Cather (CC). However, the safety criteria for Gen-IV demands demonstration of safe mitigation of whole core accident and accordingly the CC design shall consider in-vessel retention and long-term cooling of the degraded core. Whole core retention would impose considerably higher heat flux and the CC need to withstand higher thermomechanical loads. To fulfil this requirement, development of an advanced core catcher has been taken up at IGCAR, India. The main objective is to develop and qualify a refractory protective layer for the CC, which is compatible with sodium and can withstand severe thermal transients expected during corium relocation.

Based on several tests in-house, refractory magnesia was identified as a candidate material for protective lining on a stainless-steel substrate. Dedicated experiments were conducted with magnesia test specimens to study i) long-term sodium compatibility, and ii) resistance to thermal shock under simulated accident conditions. Based on the microstructure and phase analysis, the sodium compatibility was assessed whereas degradation of the specimens was determined from the destructive/ non-destructive tests before and after the experiments. Results showed the magnesia specimens to have excellent sodium compatibility and good resistance to thermal shock, indicating the magnesia lined CC to be a potential option as advanced CC for future SFRs. Design concept, experimental methods and important results are discussed in the paper.

Based on both the internal-trigger and external-trigger experiments conducted by Shanghai University of Electric Power, air entrainment is proved to be a significant factor that affects the triggering on the surface of molten droplets during fuel-coolant interaction (FCI). In this study, based on the Rayleigh equation, the mass ratio of steam to entrained air, and the disturbance amplitude and interface pressure difference at the vapor-liquid interface under different working conditions are calculated. The relationship between the air volume, the disturbance amplitude and the trigger time of the molten droplet, and the relationship between the interface pressure difference and the trigger strength of the molten droplet surface are thus analyzed. It is revealed that the air entrainment can stir the disturbance amplitude, thereby reducing the trigger time of the molten droplet. The variation of the trigger strength of the molten droplet surface is consistent with that of the vapor-liquid interface pressure difference. In order to further exhibit and verify the phenomenon, the breakup of the steam envelope with air entrainment is simulated by Moving Particle Semi-Implicit (MPS) method, and it is quantitatively estimated that under the air mass ratio of 50% and 90%, the air entrainment may reduce the trigger time of the steam envelope by nearly 1.4ms.