NDLI: Assessment of ICARE/CATHARE V1 Severe Accident Code

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Chatelard, Patrick Fleurot, Joe¨lle Marchand, Olivier Drai, Patrick
Copyright Year	2006
Abstract	The ICARE/CATHARE code system has been developed by the French “Institut de Radioprotection et de Suˆrete´ Nucle´aire” (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the “International ICARE/CATHARE Users’ Club”, under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a “circuit” experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the standalone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version (“Users’ Guidelines”). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head.
Sponsorship	Nuclear Engineering Division
Starting Page	145
Ending Page	154
Page Count	10
File Format	PDF
ISBN	0791842460
DOI	10.1115/ICONE14-89307
e-ISBN	0791837831
Volume Number	Volume 5: Safety and Security; Low Level Waste Management, Decontamination and Decommissioning; Nuclear Industry Forum
Conference Proceedings	14th International Conference on Nuclear Engineering
Language	English
Publisher Date	2006-07-17
Publisher Place	Miami, Florida, USA
Access Restriction	Subscribed
Subject Keyword	Probabilistic safety assessment Simulation Circuits Hydraulics Thermal hydraulics Pressurized water reactors Accidents
Content Type	Text
Resource Type	Article

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