Shielding options for the ITER conceptual design

Several shield options for minimizing the nuclear responses in the toroidal-field coils were analyzed for the International Tokamak Experimental Reactor (ITER) conceptual design. The total nuclear heating in the physics phase and the insulator dose in the technology phase are the most critical parameters in the design process. The first shield option has type-316 stainless steel and water shield material. Steel and water also serve as structural material and coolant, respectively. The second option is similar to the first but uses borated water instead of ordinary water. Two other options include a small layer of lead or boron carbide (B₄C) at the back of the shield. The latter three shield options were considered to reduce the nuclear heating in the toroidal-field coils relative to the steel/water shield option. An optimization process was performed, taking into consideration the thermal-hydraulics and the engineering requirements to define the shield configuration. A careful integration was performed to calculate the total nuclear heating in the toroidal-field coils which account for the neutron wall loading distribution, the change in the shield thickness in the poloidal direction, and the space between the toroidal-field coils in the divertor zone. The results show that the steel/water/Pb and the steel/borated-water shield options are very close in terms of the total nuclear heating in the toroidal field coils and the dose in the insulator material. The steel/water and steel/water/B₄C options deposit more nuclear heating in the toroidal field coils