Within the framework of lumped parameter models for integral codes, this paper focus on the modelling of a two-phase Stefan fusion problem with natural convection in the liquid phase. In particular, this specific Stefan problem is of interest when studying corium pool behavior in the framework of light water reactor severe accident analysis. The objective of this research is to analyze the applicability of different approximations related to the modelling of the solid phase in terms of boundary heat flux closure relations. Three different approximations are considered: a quadratic profile based model, a model where a parameter controls the power partitioning at the interface and the steady state conduction assumption. These models are compared with an accurate fronttracking solution of this plane fusion front problem. This "reference" is obtained by combining the same integral conservation equations as the approximate models with a mesh-based solution of the 1D heat equation. Numerical results are discussed for a typical configuration of interest for corium pool analysis. Different fusion transients (constructed from nondimensionalization considerations in terms of Biot and Stefan numbers) are used in order to highlight the potential and limitations of the different approximations.

No abstract

No abstract

In line with Polish national activities and research programs investigating non-electrical-reactor use, the national GOSPOSTRATEG-HTR project was launched, aiming at the development of a novel pre-conceptual design of a High-Temperature Gas-cooled Reactor (HTGR). The 40 MWth research reactor would serve as a technology demonstrator for future industrial purposes. In the paper, the proposal of an established thermal-hydraulic and neutronic core design is presented as a result of the National Centre for Nuclear Research team studies, in the scope of the project, including the areas of fluid mechanics, heat exchange and reactor neutronic core design support analyses. The undertaken analyses were confirmed by the series of code investigations involving integral thermal-hydraulic (MELCOR (Sandia National Laboratories, USA), CATHARE (CEA, France)), neutronic (Serpent (VTT, Finland), MCB (AGH University's Department of Nuclear Energy, Poland)), Computational Fluid Dynamics (ANSYS Fluent (ANSYS, USA)) and others. The calculations performed within the preliminary safety analysis on the pre-concept showed its compliance with international safety standards for the normal operation and Design Basis Accident sequences.