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• Two PWR1000-Like input deck were developed for MELCOR 2.2 and ASTEC V2.2 codes. • The severe accident analysis is focused on containment response ex vessel phenomena and FP release. • A SBLOCA and CSBO scenarios were simulated for 10 days. • The crosswalk results are compared, discussed and the difference highlighted. The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs). This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF). The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability. The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA). For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated. The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS). In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated. The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results. [ABSTRACT FROM AUTHOR]

• Two PWR1000-Like input deck were developed for MELCOR 2.2 and ASTEC V2.2 codes. • The severe accident analysis is focused on the RCS thermal–hydraulic and in-vessel phenomena. • SBLOCA and CSBO scenarios up to the lower head failure were simulated. • The crosswalk results are compared, discussed and differences highlighted. The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are used in Tractebel to perform source term assessments and plant Thermal Hydraulic (TH) analyses in case of Severe Accident (SA) for the Belgian Nuclear Power Plants (NPPs). These system codes were developed through intense and extensive validation activities which were mainly based on the verification of the capability of implemented physical models to reproduce the experimental results of Separate Effect Tests (SETs) and Coupled Effect Tests (CETs). However, the SETs, and in less extension the CETs, do not take into account interactions and possible synergistic or antagonistic effects which can arise among the different physical and chemical processes occurring during a SA. Thus, if a code is able to reproduce correctly most of SETs or CETs, this does not mean that it can ensure the same quality of results for the more realistic and complex Integral Effect Tests (IETs). Nevertheless, some contraindications exist also for the IETs, because of the scaling factor effect which, being difficult to quantify, could lead to development of models replicating correctly the experimental data but that are not totally reliable for reactor scale applications. In addition, waiting for the dismantling of Fukushima Dai-ichi units 1–2 and 3, which can provide new information in the field of the core degradation phenomena, in the present only data on in-vessel core melt progression for the Three Mile Island, Unit 2 (TMI-2) accident are available. Therefore, apart from simulating the TMI-2 accident scenario, there is no other way to assess models of in-vessel phenomena (H 2 production, corium pool/ debris formation and relocation, etc) implemented in the codes on reactor scale. As a consequence, there are still uncertainties in the capability of these integral tools to predict the progression of postulated severe accident transients in a NPP, notably core degradation and ex-vessel phenomena. In order to investigate the differences among the physical models adopted in MELCOR 2.2 and ASTEC V2.2 codes, and evaluate the impact that they could have on the assessment of Severe Accident Management Strategies, a crosswalk study was carried-out. This paper describes the first phase of this work, which consists of a detailed comparison between these two SA tools on a well-defined plant (PWR1000-Like) with prescribed boundary and initial conditions. Furthermore, to avoid additional and unwanted sources of discrepancies between code calculations, the two reactor models were developed in parallel following the recommendations suggested by developers for core and Reactor Coolant System (RCS) nodalization. The comparative study is focused on the TH of the RCS and in-vessel phenomena, for two different accident scenarios (SBLOCA and CSBO) up to the moment of lower-head failure. Ex-vessel behaviour will be examined in the second phase of this study. The crosswalk results obtained have shown some differences and similarities on reproducing TH in the RCS and In-vessel phenomena. Especially for these latter, the discrepancies predominate mainly due to how the codes treat corium behaviour, while thanks to the harmonisation of the initial steady-state and boundary conditions, the discrepancies on the prediction of the RCS behaviour have been minimized, at least before significant core degradation takes place. [ABSTRACT FROM AUTHOR]