One of the key components regarding heat transfer and tritium inventories in deuterium-tritium nuclear fusion reactors is the (tritium) Breeding Blanket, called Test Blanket Module or TBM in ITER experiment. Several designs are going to be tested in ITER, one of those is the HCLL (Helium Cooled Lithium Lead) design. Before being tested, it is of major interest to predict in detail several flow ...

The ARIES team has just completed the detailed design of ARIES-ACT-1 with aggressive physics and advanced SiC technology. Accurate neutronic characterizations of ARIES-ACT-1 are essential in determining its operational nuclear parameters. The neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating distribution are parameters that must be determined for tritium self-sustain...

Significant design efforts were undertaken during the Pre-Concept Design (PCD) phase of European DEMO program to optimize helium cooled pebble bed (HCPB) breeding blanket. A gate review was conducted for entire at conclusion PCD phase. This article presents a summary evolution and rationale behind HCPB blanket concept DEMO. The main performance metrics, including nuclear, thermal hydraulics, me...

Abstract One of the main functions DEMO Breeding Blanket (BB) system is to ensure tritium breeding inside reactor. Tritium a beta emitter radioactive isotope, subjected several processes that determine its permeation across materials and leakage towards environment, posing potential safety issues in terms radiological hazard. Thus, evaluation inventories components losses environment plays key ...

Accurate neutronic characterizations of fusion power plant designs are essential in determining their operational parameters. The neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating distribution are parameters that must be determined for adequate shielding and protection and fuel self-sustainability of a fusion power plant. After determining the NWL for the first wall ...

A conceptually simple first wall (FW) and blanket design for an inertial fusion system based on utilizing mobile FW tiles is presented. Using these tiles that are periodically removed, annealed, and reinstalled, tritium retention and surface erosion concerns for inertial fusion FW could be mitigated. A conceptual configuration has been developed with consideration for laser beam port accommodat...

A blanket concept made of the low electrical conductivity SiCf/SiC composite and utilizing Li17Pb83 as coolant and tritium breeder has been developed and integrated with the magnetic diversion system. Neutronics issues related to tritium breeding adequacy particularly with the area lost to the dump plates at the ring and point cusps were addressed. Radiation damage and lifetime considerations f...

Detailed three-dimensional neutronics analysis has been performed for ITER with the U.S. solid breeder blanket utilizing sintered product materials for both the breeder and multiplier. The overall tritium breeding ratio is 0.81 with two outboard breeder zones and 0.89 for the three outboard breeder zone design option. The total reactor thermal power is 1150 MW in the technology phase.