MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part I: Analysis of RCS thermal-hydraulics and in-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation Leakages Consequences Releases (MELCOR) system codes are used in Tractebel to perform source term assessments plant Thermal Hydraulic (TH) analyses case Severe (SA) for the Belgian Nuclear Power Plants (NPPs). These were developed through intense extensive validation activities which mainly based on verification capability implemented physical models reproduce experimental results Separate Effect Tests (SETs) Coupled (CETs). However, SETs, less extension CETs, do not take into account interactions possible synergistic or antagonistic effects can arise among different chemical processes occurring during a SA. Thus, if code is able correctly most SETs this does mean that it ensure same quality more realistic complex Integral (IETs). Nevertheless, some contraindications exist also IETs, because scaling factor effect which, being difficult quantify, could lead development replicating data but totally reliable reactor scale applications. In addition, waiting dismantling Fukushima Dai-ichi units 1–2 3, provide new information field core degradation phenomena, present only in-vessel melt progression Three Mile Island, Unit 2 (TMI-2) accident available. Therefore, apart from simulating TMI-2 scenario, there no other way assess phenomena (H2 production, corium pool/ debris formation relocation, etc) scale. As consequence, still uncertainties these integral tools predict postulated severe transients NPP, notably ex-vessel phenomena. order investigate differences adopted MELCOR 2.2 ASTEC V2.2 codes, evaluate impact they have assessment Management Strategies, crosswalk study was carried-out. This paper describes first phase work, consists detailed comparison between two SA well-defined (PWR1000-Like) with prescribed boundary initial conditions. Furthermore, avoid additional unwanted sources discrepancies calculations, parallel following recommendations suggested by developers Reactor Coolant System (RCS) nodalization. comparative focused TH RCS scenarios (SBLOCA CSBO) up moment lower-head failure. Ex-vessel behaviour will be examined second study. obtained shown similarities reproducing In-vessel Especially latter, predominate due how treat behaviour, while thanks harmonisation steady-state conditions, prediction been minimized, at least before significant takes place.