ETD

• melcor
• molten core concrete interaction
• uncertainty and sensitivity analysis

Severe Accidents (SA) are low-probability but high-consequence events in which the failure of multiple safety systems can lead to the loss of core cooling and, in the most critical scenarios, to the partial or complete meltdown of nuclear fuel. In such conditions, molten core material (corium) may relocate to the lower head of the reactor pressure vessel (RPV), potentially causing its failure. Once the vessel fails, the corium slumps into the cavity located beneath the RPV and comes into contact with the concrete basemat of the containment structure, leading to a Molten Core – Concrete Interaction (MCCI).
To analyse these complex scenarios, integral codes, such as MELCOR, are widely used to simulate the evolution of severe accidents. Nevertheless, the reliability of these simulations is strongly affected by uncertainties, which may originate from physical modelling assumptions, numerical approximations, and the lack of precise knowledge of boundary conditions. In this framework, the Uncertainty and Sensitivity Analysis (UaSA) approach has emerged as a fundamental methodology, as it allows combining Best Estimate simulations with a systematic quantification of uncertainties, thus providing a more comprehensive and robust basis for safety assessments.
The objective of this thesis is to assess the influence of corium properties uncertainties, such as viscosity, conductivity, density, emissivity and specific heat, on key Figures of Merit (FoMs), including concrete ablation, containment pressure and gas flammability during the ex-vessel phase of a severe accident.
These FoMs are chosen because they represent the most safety-relevant outcomes of a severe accident involving corium – concrete interaction. Concrete ablation directly quantifies the progression of the molten core through the basemat, indicating the potential for containment bypass or basemat melt-through (epsilon mode of containment failure). Containment pressure reflects the global thermodynamic response of the containment system, which is critical for assessing structural integrity and the likelihood of over-pressurization failures (delta mode). Gas flammability, related to the generation and accumulation of hydrogen and carbon monoxide, is a key parameter for evaluating the risk of combustion phenomena that could challenge containment integrity (gamma mode). Together, these FoMs capture the main physical hazards arising from corium behaviour and their coupling with containment performance, thereby providing a comprehensive basis for uncertainty and sensitivity analysis.
The first step of the analysis consists in identifying a suitable range for each corium property. To this end, a literature review was carried out, considering two reference corium compositions: a fully oxidic one and a partially metallic one. The selected severe accident scenario is a Station Blackout (SBO).
In the first part of the thesis, a 1000 MWe PWR reference design is considered, with two alternative cavity configurations: dry cavity, in which no water is injected onto the debris pool and wet cavity, in which 200 tons of water are introduced, fully covering the debris pool and filling the cavity.
In the second part, a Fukushima – like scenario is analysed for a BWR reference plant, considering only the dry cavity configuration.
For both reactor designs, a stand-alone containment model is adopted. This modelling approach is easier to handle and significantly reduces computational time, as the simulation is initiated directly from the beginning of the ex – vessel phase. For this purpose, an instantaneous melt-spreading is considered, in which the debris pool completely fills the cavity at the start of the transient. Initial and boundary conditions are imported from the corresponding full – plant MELCOR model.