This report presents the activity performed in the framework of LP1, Objective B (Accident consequent evaluation) task B1 Topic 3 (Integral study of accident sequences with reference to BWR and PWR reactor next to Italian borders) of PAR 2014, ADP ENEA-MSE. Three different PWR accidents - Short term Station Blackout (SBO), Loss of Feed Water (LFW), Large Break Loss of Coolant Accident (LBLOCA) - scenarios have been studied by using the MELCOR code. To lead these three accidents to a BDBA, they will be unmitigated and the actions of the operator are assumed to fail; the characterization of the thermal hydraulic behaviour, the in-vessel phenomena, the core degradation and corium behaviour in the lower head are here analyzed. A first estimation of the source term is here also presented for the SBO scenario. For the BWR, a preliminary Fukushima Unit 1 like MELCOR nodalization has been deveped allowes to have a basis to analyse transients scenarios and compare the data with calculated results of other codes and possible full scale plant transient data, if available. The analyses of thermal hydraulic phenomenology of interest, for the validation of severe accident code, related to advanced reactor (as Small Modular Reactor) that could be in operation in the next short term, are here presented. The “scaling issue”, that determines uncertainty in the code calculated data, is here briefly analyzed considering the aspect of interest that should be investigated for the “severe accident” analyses.

Integral study of accident sequences with reference to NPPs next to the Italian borders

Mascari, Fulvio
2015-09-25

Abstract

This report presents the activity performed in the framework of LP1, Objective B (Accident consequent evaluation) task B1 Topic 3 (Integral study of accident sequences with reference to BWR and PWR reactor next to Italian borders) of PAR 2014, ADP ENEA-MSE. Three different PWR accidents - Short term Station Blackout (SBO), Loss of Feed Water (LFW), Large Break Loss of Coolant Accident (LBLOCA) - scenarios have been studied by using the MELCOR code. To lead these three accidents to a BDBA, they will be unmitigated and the actions of the operator are assumed to fail; the characterization of the thermal hydraulic behaviour, the in-vessel phenomena, the core degradation and corium behaviour in the lower head are here analyzed. A first estimation of the source term is here also presented for the SBO scenario. For the BWR, a preliminary Fukushima Unit 1 like MELCOR nodalization has been deveped allowes to have a basis to analyse transients scenarios and compare the data with calculated results of other codes and possible full scale plant transient data, if available. The analyses of thermal hydraulic phenomenology of interest, for the validation of severe accident code, related to advanced reactor (as Small Modular Reactor) that could be in operation in the next short term, are here presented. The “scaling issue”, that determines uncertainty in the code calculated data, is here briefly analyzed considering the aspect of interest that should be investigated for the “severe accident” analyses.
Analisi incidentale;Rapporto tecnico;Scaling;Reattori nucleari avanzati;Incidenti severi;Sicurezza nucleare
File in questo prodotto:
File Dimensione Formato  
ADPFISS-LP1-059.pdf

accesso aperto

Licenza: Creative commons
Dimensione 2.94 MB
Formato Adobe PDF
2.94 MB Adobe PDF Visualizza/Apri

I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.

Utilizza questo identificativo per citare o creare un link a questo documento: http://hdl.handle.net/20.500.12079/7832
Citazioni
  • ???jsp.display-item.citation.pmc??? ND
  • Scopus ND
social impact