In-vessel Loss Of Coolant Accident (LOCA) is one of the Design Basis Accident to be considered to support the future DEMOnstration power plant safety assessment. The water-cooled lithium-lead (WCLL) Breeding Blanket (BB) concept relies on Lithium-Lead as breeder, neutron multiplier and tritium carrier. The breeding modules are cooled by two independent pressurized water systems: the fist-wall (FW) and the breeding zone (BZ) coolant systems. The postulated initiating event (PIE) considered for this safety analysis is a double ended pipe rupture of the blanket module first wall channels. This event causes the inlet of coolant into the plasma chamber volume triggering an unmitigated plasma disruption and the pressurization of the Vacuum Vessel (VV) volume. The fusion version of MELCOR code (ver. 1.8.6) is used to evaluate accident consequences for two different scenarios, with the presence and absence of the downstream isolation valves, respectively. The chemical reaction between the coolant and the first wall tungsten layer inside the VV has been considered together with the mobilization of the radioactive source term. Pressure and temperature transient behavior in the tokamak volumes demonstrate that safety margins are respected during the accidental sequence.
Preliminary safety analysis of an in-vessel LOCA for the EU-DEMO WCLL blanket concept
Porfiri M. T.;
2020-01-01
Abstract
In-vessel Loss Of Coolant Accident (LOCA) is one of the Design Basis Accident to be considered to support the future DEMOnstration power plant safety assessment. The water-cooled lithium-lead (WCLL) Breeding Blanket (BB) concept relies on Lithium-Lead as breeder, neutron multiplier and tritium carrier. The breeding modules are cooled by two independent pressurized water systems: the fist-wall (FW) and the breeding zone (BZ) coolant systems. The postulated initiating event (PIE) considered for this safety analysis is a double ended pipe rupture of the blanket module first wall channels. This event causes the inlet of coolant into the plasma chamber volume triggering an unmitigated plasma disruption and the pressurization of the Vacuum Vessel (VV) volume. The fusion version of MELCOR code (ver. 1.8.6) is used to evaluate accident consequences for two different scenarios, with the presence and absence of the downstream isolation valves, respectively. The chemical reaction between the coolant and the first wall tungsten layer inside the VV has been considered together with the mobilization of the radioactive source term. Pressure and temperature transient behavior in the tokamak volumes demonstrate that safety margins are respected during the accidental sequence.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
