Experimental Investigation of Melt Coolability Behaviour in an Ex-Vessel Core Catcher - the Effect of Flooding Time

Some of the advanced nuclear reactors employ an ex-vessel core catcher to mitigate core melt scenarios by stabilizing and cooling the corium for prolonged period by strategically flooding it. The side indirect cooling with top flooding strategy described in this study may lead to water ingression either through the melt crust which may lead to interaction between un-oxidised metal in the melt and water leading to hydrogen production. In order to avoid this deleterious scenario, water ingression into the bulk of the melt should be avoided. The studies described in this manuscript show that water ingression depends on the flooding strategy, i.e. the time delay between top flooding and melt relocation. Two experiments under identical conditions of simulant temperature, melt material and test section geometry were conducted with simulated decay heat of 1 MW/m3. Sodium borosilicate glass was used as the corium simulant. In the first experiment, water was flooded onto the top of melt pool soon after melt relocation. In the second experiment, water flooding at the top of melt pool was made after 30 minutes of the melt relocation. The results show that a finite time delay of introduction of water onto the top of the melt pool is paramount to engender the development of a stable crust around the melt and therefore eliminating water ingression into melt pool and ensuring controlled coolability of the melt.

Numerical Simulation of an Out-Vessel Loss of Coolant from the Breeder Primary Loop Due to Large Rupture of Tubes in a Primary Heat Exchanger in the DEMO WCLL Concept

This work presents a thermohydraulic analysis of a postulated accident involving the rupture of the breeder primary cooling loop inside a heat exchanger (once through steam generator). After the detection of the loss of pressure inside the primary loop, a plasma shutdown is actuated with a consequent plasma disruption, isolation of the secondary loop, and shutoff of the pumps in the primary; no other safety counteractions are postulated. The objective of the work is to analyze the pressurization of the primary and secondary sides to show that the accidental overpressure in the two sides of the steam generators is safely accommodated. Furthermore, the effect of the plasma disruption on the FW, in terms of temperatures, should be analyzed. Lastly, the time transients of the pressures and temperatures in the HX and BB for a time span of up to 36 h should be obtained to assess the effect of the decay heat over a long period. A full nodalization of the OTSG was realized together with a simplified nodalization of the whole PHTS BB loop. The code utilized was MELCOR for fusion version 1.8.6. The accident was simulated by activating a flow path which directly connected one section of the primary with the parallel section of the secondary side. It is shown here that the pressures and the temperatures inside the whole PHTS system remain below the safety thresholds for the whole transient.

Heat transfer at the step of a CANDU calandria during a severe accident

During the in-vessel stage of a severe accident in a CANDU 6 reactor, decay heat from a collapsed core would be rejected through the calandria walls into the surrounding water. At the step in the calandria wall, the subshell and annular plate meet at a right angle pointing into the calandria. The geometry at this joint could concentrate the exiting heat flux, potentially leading to calandria failure. Finite element analysis is used to study the heat transfer near the welded joint. Different weld profiles, boundary conditions, and decay heat characteristics are considered, and the local concentration of exiting heat flux is calculated.