Development of ceramic breeding zone models and mock-up of a DEMO blanket for in-pile irradiation testing

Abstract

Abstract A goal of demonstration fusion reactor (DEMO) with ceramic helium cooled (CHC) blanket test module (BTM) is to demonstrate the breeding capability that would lead to tritium self-sufficiency in the ITER reactor and to extract a high-grade heat suitable for electricity generation. The experimental validation of all the adopted design solutions is one of main concerns at design and calculation works carried out with the aim to create the CHC BTM. The in-pile test is one of the most important components of the breeding zone feasibility validation. For validation of the CHC BTM breeding zone feasibility we have developed and fabricated two models and breeding blanket mock-up for testing in the IVV-2M reactor. The first model and mock-up contain pellets from lithium orthosilicate and porous beryllium, the second model contains pebbles from these materials. The tritium produced in the breeder material is purged by flow of neon at 0.1–0.2 MPa. The models structural material is ferrite martensite steel. A special processing installation has been designed, constructed and assembled at the IVV-2M reactor for study of the kinetics of tritium extraction from ceramics under the reactor irradiation. Designs of the models and experimental channel and results of neutronie and thermo-hydraulic calculations are presented in the paper.