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The water-cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant(DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium– lead as breeder–multiplier,pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach isemployed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by meansof radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidaldirection. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operationand selected postulated initiating events. Water and lithium–lead manifolds are designed and integrated with a consistentprimary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium–leadsystem. Rationale and features of the single-module-segment water-cooled lithium–lead breeding blanket design arediscussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration withthe primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areasof research, and development needs are finally pointed out.

The water-cooled lithium-lead breeding blanket is a candidate option for the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015 specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial-toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium-lead flows in a radial-poloidal direction. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. Water and lithium-lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium-lead system. Rationale and features of the single-module-segment water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium

Pubblicato su: PIANURA: SCIENZE E STORIA DELL'AMBIENTE PADANO vol. 13

The Steam Generator Tube Rupture postulated event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system needs to by deeply investigated due to the possible hazardous consequences affecting the structural integrity of internals. These reactor designs have steam generators inside the reactor vessel, thus, the interaction between the secondary side coolant (water) and the HLM (e.g. steam generator tube rupture) has to be considered as challenging safety issue in the design and also in the preliminary safety analysis of these reactor types. The LIFUS5/Mod2 facility experiments series B were executed in the framework of the FP7 EC Lead-cooled European Advanced Demonstration Reactor – LEADER – project. The test section was representative of the Spiral Tube Steam Generator of European Lead Fast Reactor. Water at about 180 bar and 270°C was injected into Lead Bismuth Eutectic alloy at about 2 bar and 400°C. The injection was performed in the center of a tube bundle composed of 188 tubes (a representative portion of SG bundle). This analysis is aimed at providing engineering feedbacks to SG designers (e.g. domino effect and mechanical loadings on components). Moreover, the acquired experimental data constituted a wide database for development and validation of numerical models and calculation codes. © 2019 American Nuclear Society.

In the framework of EU MAXSIMA project, the safety of the Steam Generator (SG) adopted in the primary loop of the Heavy Liquid Metal Fast Reactors (HLMFR) has been treated by looking at consequences and damage propagation of a SG Tube Rupture event (SGTR) and by characterizing leak rates from typical cracks. Indeed, instrumentation able to promptly detect the presence of a crack in the steam generator tubes may be used to prevent its further propagation which would possibly lead to a full rupture of the tube. Indeed, the application of the leak before break concept is relevant for improving the safety of a reactor system and decreasing the probability of the pipe break event. In this framework, a new experimental campaign (Test Series C) has been carried out in LIFUS5/Mod3 facility, installed at ENEA C.R. Brasimone, in order to characterize and correlate the leak rate through typical cracks occurring in the pressurized tubes with signals detected by proper transducers. Test C1.3_60 was executed injecting water at about 20 bar and 200°C into Lead Bismuth Eutectic alloy (LBE). The injection was performed through a laser micro-holed plate of 60 micrometer in diameter. The analysis of the thermo-hydraulic data permitted to characterize the leakage through typical cracks which can occur in pressurized tubes of steam generator, while the analysis of the data acquired by the microphones and accelerometers highlighted that is possible to correlate the signals to the leakage, and the rate of release.