Breeding zone models of demo ceramic helium cooled blanket test module for reactor in-pile testing

摘要

The goal of a demonstration fusion reactor ceramic helium cooled (CHC) blanket test module (BTM) is to demonstrate a breeding capability that would lead to tritium self-sufficiency in the ITER reactor and to heat with parameters suitable for electricity generation. For validation of the CHC BTM breeding zones feasibility we have developed and fabricated two models for testing in an IVV-2M reactor. Structural material of the models is ferritic-martensitic steel. The breeder material is lithium orthosilicate in pebble-beds and pellet forms. The multiplier material is beryllium in pebble-beds and porosity forms. Cooling is provided by helium at 10?MPa. The tritium produced in the breeder material is purged by a flow of helium + 0.1% of hydrogen at 0.1-0.2?MPa. Designs of models are described and results of neutronic and thermo-hydraulic calculations are presented.