Development of tungsten armored high heat flux plasma facing components for ITER like divertor application

Abstract

The dome and reflector plate are the parts of divertor plasma facing components (PFCs) of ITER tokamak which are mainly aimed for the removal of heat load of maximum 5 MW/m2 in steady state condition. The dome is a curved tungsten armoured component and the reflector plate is a straight component. These components have multi-layered joints made of various materials such as tungsten (W), OFHC copper (Cu), copper alloy (CuCrZr) and stainless steel (SS316LN). Joining of such multi-layered joints is known to be problematic due to joining of several dissimilar materials. In this paper, we report the indigenous development of medium size dome and reflector plate via vacuum brazing route for ITER like tokamak application. In order to evaluate the performance of the dome against ITER-like scenarios (maximum heat flux removal of 5MW/m2), the dome has been successfully tested for 1000 number of steady-state thermal cycles at incident heat fluxes of 3.87 MW/m2 in the High Heat Flux Test Facility (HHFTF) at IPR. Subsequent testing of additional 200 thermal cycles was also done at incident heat flux of 6 MW/m2. During the High heat flux (HHF) tests, surface temperature of W tiles reached 640oC and the beam power was restricted at 6MW/m2 to limit the temperature below 450oC at the CuCrZr heat sink. Total 1200 steady-state thermal cycles have been completed. At 6 MW/m2, the absorbed heat flux was 4 MW/m2. Engineering analysis on the HHFT of the dome has been performed using Finite element method (FEM) and Computational Fluid Dynamics (CFD) to simulate and to correlate with the experimental data. Ultrasonic immersion technique – Non destructive testing (NDT) was used to inspect the brazed joint quality of the dome before and after the HHFT. The results of the experimental details, engineering analysis and methodology adopted to fabricate the medium size dome and reflector plate are presented here.