Fusion Science and Technology / Volume 60 / Number 1 / July 2011 / Pages 264-271
In-Vessel Components - FW, Blanket, Shield & VV / Proceedings of the Nineteenth Topical Meeting on the Technology of Fusion Energy (TOFE) (Part 1)
The U.S. Dual Coolant Lead Lithium (DCLL) ITER Test Blanket Module (TBM) is under development for operation in the ITER reactor. The DCLL TBM must satisfy the Structural Design Criteria for ITER In-vessel Components (SDC-IC), which provides rules for the design evaluation and stress analyses of in-vessel mechanical components of ITER with the purpose of ensuring that required safety margins are maintained relative to the types of mechanical damage which might occur as a result of imposed loadings.
Primary stresses on the blanket structure come from the pressurization of coolants, the weight of the blanket element, and any electromagnetic forces due to plasma disruptions events. Secondary stresses in the materials due to thermal stress resulting from temperature gradients also contribute to the stress state of the structure. The response to primary stresses will depend on the distribution of loads, the blanket support, as well as material thermo-physical properties, which depend on operating temperatures, loads, fabrication and heat treatment and changes caused by neutron irradiation effects.
A detailed structural and thermal analysis of the DCLL TBM under typical loading conditions was performed. Highly stressed locations in the TBM were identified and the stress was broken down into membrane, bending, secondary, and peak stress for evaluating local stress intensities and equivalent stress in order to apply the SDC-IC design rules. Both low- and high temperature damage rules were evaluated to show lack of excessive deformation and negligible thermal creep.