A goal of demonstration fusion reactor (DEMO) with ceramic helium cooled (C
HC) blanket test module (BTM) is to demonstrate the breeding capability tha
t would lead to tritium self-sufficiency in the ITER reactor and to extract
a high-grade heat suitable for electricity generation. The experimental va
lidation of all the adopted design solutions is one of main concerns at des
ign and calculation works carried out with the aim to create the CHC BTM. T
he in-pile test is one of the most important components of the bleeding zon
e feasibility validation.
For validation of the CHC BTM breeding zone feasibility we have developed a
nd fabricated two models and breeding blanket mock-up for testing in the IV
V-2M reactor. The first model and mock-up contain pellets from lithium orth
osilicate and porous beryllium, the second model contains pebbles from thes
e 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 marten
A special processing installation has been designed, constructed and assemb
led at the IVV-2M reactor for study of the kinetics of tritium extraction f
rom ceramics under the reactor irradiation.
Designs of the models and experimental channel and results of neutronic and
thermohydraulic calculations are presented in the paper.