A Severe Accident countermeasure can be the retention of the melted core inside the reactor vessel. A way to achieve this kind of corium retention is by removing heat from the steel vessel by flooding the reactor cavity. A first calculation was performed in order to assess In-Vessel Core Damage Retention (IVR) possibilities for an AP1000 reactor as a consequence of a Beyond Design Basis Accident (BDBA). The computer code used to perform this calculation is ASTEC. ASTEC is an integral code to analyse severe accidents in LWRs, jointly developed by IRSN (France) and GRS (Germany), and it is meant to predict a whole severe accident sequence from the initiating event to source term evaluation inside the containment building. The accidental transient was carried out using ICARE and CESAR modules for the evaluation of the primary system thermal-hydraulic behaviour. The improved debris bed and magma models available in the code last version have been used to describe the late phase of core degradation. In this calculation different assumptions were done regarding the corium, particularly different configurations of the corium relocated in the RPV lower head were considered. The results of the calculations show the effectiveness of the reactor cavity flooding option as a severe accident management strategy.
In Vessel Retention Analisys with Astec code
DI GIULI, MIRCO;SUMINI, MARCO;ROSSI, FABIANA;
2012
Abstract
A Severe Accident countermeasure can be the retention of the melted core inside the reactor vessel. A way to achieve this kind of corium retention is by removing heat from the steel vessel by flooding the reactor cavity. A first calculation was performed in order to assess In-Vessel Core Damage Retention (IVR) possibilities for an AP1000 reactor as a consequence of a Beyond Design Basis Accident (BDBA). The computer code used to perform this calculation is ASTEC. ASTEC is an integral code to analyse severe accidents in LWRs, jointly developed by IRSN (France) and GRS (Germany), and it is meant to predict a whole severe accident sequence from the initiating event to source term evaluation inside the containment building. The accidental transient was carried out using ICARE and CESAR modules for the evaluation of the primary system thermal-hydraulic behaviour. The improved debris bed and magma models available in the code last version have been used to describe the late phase of core degradation. In this calculation different assumptions were done regarding the corium, particularly different configurations of the corium relocated in the RPV lower head were considered. The results of the calculations show the effectiveness of the reactor cavity flooding option as a severe accident management strategy.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
