Progress on PWR lower head failure predictive models

Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head

“…The OLHF1 experiment, performed by SNL (USA), intended to simulate the thermal/mechanical loads to a 1/4.85 -scale model of a reactor pressure vessel. This experiment was already included in V1.3 version validation matrix (Koundy et al, 2008) (Bandini et al, 2010). The same modelling was considered for V2.0 calculations.…”

“…14. These calculations were performed using the 'OEUF' vessel rupture model (Koundy et al, 2008) with the criterion for the vessel rupture corresponding to excessive creep deformation (CRIT = 6). Note that the excessive creep deformation criterion was the one used by default in V1.3 version, whereas all criteria are now checked by default in V2.0 version (CRIT = 0 means that fragile and ductile failure, wall thickness and displacement, creep velocity and deformation are all considered, see Table 3).…”

In-Vessel Core Degradation

“…Based on these properties, extensive numerical analyses were done for the experiments [64] [65] [94]. Different creep and damage models were applied to predict the vessel failure (ranging from a simple Norton-Bailey approach to the more sophisticated Lemaitre-Chaboche approach).…”

Unstable Crack Propagation Under Severe Accident Scenario Conditions in a Pressurized Water Reactor

“…The tests led to the same conclusions about the material behaviour variability as the LHF-OLHF tests for the French material. Some simplified numerical analysis were made in order to simulate these tests [7][8][9][10]. From this established fact, a lot of tests were carried out in order to characterize the French material behaviour (16MND5) according to its steel grades [11].…”