Within the EU-SAFEST project, the Swedish Radiation Safety Authority (SSM) proposed to perform tests on molten corium-concrete interaction (MCCI) with basaltic concrete and BWR-specific corium, characterized by a higher Zr/U ratio than PWR corium. These tests (denoted SSM-1a/2b/2c/3) were carried out in Framatome's SICOPS facility in Erlangen. Once the melt initially containing ≈75 wt% UO2, ≈17 wt% ZrO2, was generated and its interaction with the concrete had started, metallic Zr was subsequently added to the melt from the top. All tests showed a fast concrete ablation by the melt, faster than in previous SICOPS experiments without Zr addition, most likely due to the higher temperature of the melt caused by the Zr oxidation. The absence of crusts and even erosion profiles at the bottom seen during post-test examination support the assumption of a homogeneously mixed melt pool during MCCI. In the presence of high melt temperatures and the vivid mixing by rising concrete decomposition gases, local crusts between melt and concrete should not be stable and no evidence for them has been found. As the formation of such crusts is seen as the main potential mechanism for anisotropic concrete ablation in the early stages of an MCCI, the investigated conditions with Zr being present in the melt can be considered capable of suppressing this effect. This is important since the core melt, after its release from the RPV, typically contains a high fraction of non-oxidized Zr. While no crusts were found at the interface, sampling in the late phase of the SSM-3 test revealed evidence for the formation of a kind of transition zone between melt and concrete. This conclusion was drawn based on tactile feedback during sampling from the bottom of the melt pool where a soft/viscous zone could be felt and from the appearance of taken samples including small solid aggregates from the transition zone. Gas measurements during SSM-3 showed a significant H2 production, due to the reaction of water from the concrete with metallic zirconium. The peaks in the H2 concentration occurred when the concrete erosion velocities were highest. This confirms that, even at high erosion and superficial gas rates, the Zr in the bulk is still capable of reducing the percolating steam.

The steam explosion plays an essential role in the safety analysis of light water reactors (LWRs). Some studies have demonstrated that the occurrence of steam explosions is dependent on many factors such as melt and coolant temperatures, melt and coolant properties, non -condensable gases, etc. After the Fukushima accident, seawater as an emergency coolant and its impact on fuel coolant interactions are receiving attention. However, there is still little knowledge on the impact of seawater on steam explosion. The present study is intended to examine the effect of coolant salinity on steam explosion through a series of tests with single molten droplet falling in different coolant pools (DI water, and seawater at different salinities from 7.7 g/kg to 35 g/kg). The experimental results reveal that the salinity of coolant significantly influences the probability of spontaneous steam explosion of molten tin droplets. The probability of steam explosion generally increases with increasing salinity from 0 to 17.5 g/kg. The molten droplet in seawater experiences more pronounced deformation at same depth before the vapor film of the droplet collapses. What's more, the peak pressure generated by steam explosion in seawater is notably higher than that in DI water. The fragmentation of molten tin droplet after the explosion is enhanced accordingly.

Melt infiltration into porous media is an intriguing phenomenon that holds immense significance across various sciences and technologies. In this work, the problem of metallic melt infiltration in particulate beds is investigated for understanding and prediction of severe accident progression associated with a molten pool penetrating through an underlying debris bed which may form in the reactor core or in the lower head of a light water reactor. The present study aims to quantify the effect of particle surface's wettability on melt infiltration kinetics. For this purpose, two categories of experiment are conceived and carried out to measure the wettability of different material surfaces by melt and to characterize melt infiltration kinetics in one-dimensional particulate beds, respectively. The melt material is tin–bismuth eutectic alloy with a melting point of 139 °C. Copper (Cu), stainless steel (SS), Tin (Sn) and tin-coated stainless steel (Sn-coated SS) are chosen as materials of substrates and particles in wettability measurement and melt infiltration study. The particulate beds, packed with 1.5 mm spheres, are preheated to 200 °C before the melt infiltration begins. The experimental data of wettability measurement shows that the contact angles of liquid Sn-Bi eutectic on the above-mentioned material surfaces range from 79° to 135°. The results of melt infiltration tests confirm the significant effect of wettability on melt penetration kinetics. The capillary force plays a significant role in the initial infiltration of particulate beds. Specifically, a wettable particulate bed enhances the initial melt infiltration, whereas non-wettable beds hinder it.

In the context of severe accidents, considerable research efforts throughout the world are currently directed towards ex-vessel corium behavior. The NEA Reduction of Severe Accident Uncertainties (ROSAU) project aims to reduce knowledge gaps and uncertainties associated with two areas: the spreading of core melt in the containment cavity as well as ex-vessel core melt and debris coolability. One pre-test and five large underwater melt spread tests (MST) with molten prototypic material in a newly designed facility are conducted at the Argonne National Laboratory (ANL) in the United States, under the co-ordination with the US Nuclear Regulatory Commission. Part of KTH contributions is to provide numerical results with the developed Moving Particle Semi-implicit (MPS) method code, with specific focus on the temperature distribution and leading-edge progression. The MST-0 and MST-2, conducted in a dry and wet spreading channel, are simulated in the present study. The predicted temperature by MPS code indicates a noticeable decrease at the melt leading edge and a slow decrease in bulk melt in both simulations. Additionally, it is found that the MPS code underestimates the melt average thickness in both simulations due to the absence of a debris porosity model. Overall, the simulation results suggest that the MPS code predicts the melt leading-edge progression and immobilization for all the tests.

A numerical model to simulate molten metal relocation with phase change is proposed, coupling the level-set method to track the metal-gas interface and an enthalpy-porosity model to handle phase changes between solid and liquid metal. This coupling simultaneously solves the evolution of the metal-gas interface and liquid-solid metal. The numerical model is validated by a melting experiment involving a Sn–Bi eutectic alloy on a copper substrate, wherein the alloy's transient morphology and spreading diameter are measured. The numerical simulation effectively replicates the observed melting and spreading behaviors of the metal on the solid surface. Further validations, including a melt infiltration simulation and experiment, are consistent with findings from previous research. These simulations affirm the model's capability and efficiency in accurately representing the dynamics of melt relocation across various geometries, even within complex porous structures.

During a severe accident in light water reactors, the molten reactor core (corium) falls into a water pool in the form of a jet. Complex interactions may occur between the melt and coolant known as molten fuel-coolant interactions (FCI), including energetic coolant evaporation and metallic melt (e.g., Zr and Fe) oxidation. This may further lead to steam and hydrogen explosions, which are both substantial safety risks for nuclear power plants. The heat of reaction and hydrogen production during oxidation can influence the progress and severity of the accidents. For example, the reaction heat may prolong the liquid state of corium, potentially leading to highintensity explosions, whereas the generated hydrogen can create a combustible atmosphere, increasing the risk of hydrogen explosion. Therefore, this study evaluates the hydrogen production and oxidation degree of molten metallic droplets falling into a water pool to improve the FCI models for the risk evaluation of severe accident safety. The MISTEE-OX facility at KTH, which has been primarily built to study steam explosions is modified to investigate oxidation during FCI and provide experimental data on the oxidation behaviour of metallic droplets (Zr/Fe) quenched in a subcooled water pool. The dynamics of the falling droplets and generated bubbles are recorded using a high-speed camera, and the total volume of the bubbles is measured using a graduated cylinder. This study presents preliminary experimental results of the oxidation between Zr/Fe droplets and water, as well as recent improvements in measurement methods and facility upgrades. Our research findings are useful to enhance the knowledge of the oxidation process in FCI phenomena and validate the related mechanistic models in FCI codes.

Motivated by a need to characterise debris remelting phenomena which may occur during the progression of a severe accident in light water reactors, experimental studies on melt penetration in a debris bed have been carried out at KTH. To help understand experimental observations and obtain more detailed information of melt infiltration inside debris beds, a numerical study on melt penetration in particulate beds is presented in this paper. The Level set method was adopted through the COMSOL Multi-physics platform to track the melt-gas multiphase flow in particulate beds. The numerical model is primarily validated against available experiments. Further simulation results show the bed's wettability significantly affects the dynamics of melt penetration in a preheated particulate bed when the capillary force is relatively higher than the inertial force. In addition, melt initially penetrates deeper and faster in wettable particulate beds.

The generic severe accident management guidelines (SAMG) were developed as a response to TMI-2 accident to improve the defense-in-depth (DiD) concept of light water reactors (LWRs). Various SAMGs were developed for plant specific application considering different objectives of severe accident management (SAM) and design features of nuclear reactors. To verify and validate the effectiveness of SAMG and SAM actions for mitigating accident consequences and terminating accident progression, the analytical simulation through best estimate codes were performed extensively to provide quantitative details for the assessment of an SAMG and its actions. The present study is carried out to review the representative works concerned with the assessment of SAMG actions in the pressurized water reactors (PWR) and European VVERs using analytical simulation. The outcomes would be valid to realize the improvement and development of assessment methodology in future studies.

Analytical simulation using best-estimate codes were suggested to be extended for the elaboration and improvement of SAMG in current PWRs. This work is about to perform an assessment work using MELCOR for the effectiveness of safety injection to achieve the PBF strategy in SAM following a BDBA scenario, that is LOCA with concurrent SBO. In the simulations, the safety injection is assumed to be retrieved with the postulated power recovery at different timing during core relocation. The simulation results illustrates that the grace period of preventing vessel failure varies with LOCA break size and locations. The safety injection implemented in grace period is capable of retarding or ceasing the core relocation, sequentially avoiding the massive core relocation into lower plenum, mitigating the hydrogen generation and fission product release from core. Meanwhile, the injection later than grace period would be failed to prevent RPV failure, and it negatively affects hydrogen generation in some scenarios. The results also indicate that the smallest injection capacity of HPSI system in Swedish PWR is sufficient to achieve the effective mitigation.

The experiment QUENCH-20 with BWR simulation bundle was conducted at KIT on 9th October 2019. This test was performed in the framework of EU SAFEST project and its international access to European SA research infrastructure of the users from Swedish Radiation Safety Authority (SSM) in collaboration with Westinghouse Sweden, GRS and KTH. The test objective was the investigation of the BWR fuel assembly degradation including a control blade with B4C neutron absorbers. The test bundle represented one quarter of a BWR fuel assembly. 24 electrically heated fuel rod simulators were filled with krypton (pressure of about 0.6 MPa), whereas the holes of absorber pins were filled with helium (overpressure 0.02 MPa). According to the pre-test calculations performed with ATHLET-CD, the bundle was heated to a temperature of 1230 K at the cladding of the central rod at the hottest elevation of 950 mm. This pre-oxidation phase in steam lasted 4 h. During the following transient stage, the bundle was heated to a maximal temperature of 2000 K. The cladding failures were observed at temperatures about 1700 K and lasted about 200 s. Massive absorber melt relocation was observed 50 s before the end of the transient stage. The test was terminated by the injection of quench water with a flow rate of 50 g/s into the bundle bottom. Fast temperature escalation from 2000 to 2300 K during 20 s was observed. The mass spec-trometer measured release of COx and little CH4 during the reflood as products of absorber oxidation; corre-sponding mass of reacted B4C was 4.3% of the total mass of B4C pins. Hydrogen production during the reflood amounted to 32 g (57.4 g during the whole test) including 10 g from B4C oxidation. The results of the post-test simulations with the AC2/ATHLET-CD code show a good agreement with the experimental observations con-cerning the thermal behavior. However, melting and relocation of BWR components was not calculated and the B4C oxidation was underestimated.