This benchmark is based on a well-defined problem concerning a pressurised water reactor (PWR) main steam line break, which may occur as a consequence of the rupture of one steam line upstream of the main steam isolation valves. This event is characterised by significant space-time effects in the core caused by asymmetric cooling and an assumed stuck-out control rod during reactor trip. It is based on reference design and data from the Three Mile Island Unit 1 Nuclear Power Plant (TMI-1). It includes a description of the event sequence with set points of all activated system functions and typical plant conditions during the transient.
This report summarises the results contributed by international participants concerning Phase II of the exercise: a coupled 3-D neutronics/core thermal-hydraulics response evaluation using inlet and outlet core transient boundary conditions.
Classification ID: NEA/NSC/DOC(2002)12