L Giancarli et al. Fusion Engineering and Design 61-62(2002)307-318 3.1.3. Analysis Detailed analyses of the ARIES-at blanket were performed and the results are summarized are summarized below [1I A tritium-breeding ratio of 1. I was calculated from 3D neutronics analyses of the power core. Thermal-hydraulic analyses conservatively as- MHD-laminarized Pb-17Li flo showed that for an average outlet Pb-17Li emperature of 1100C, both the maximum SiC!Sic temperature at the FW and the max imum blanket SiC/Pb-17Li interface tempera ture at the inner channel wall are maintained at 1000C, which satisfy the maximum tem perature limits shown in Table 1. The corre sponding blanket pressure drop is about 0.25 Fig. 1. ARIES-AT outboard first wall and blanket segment. Stress analyses were performed both on the module outer and inner shells indicating that Pb-17Li at a high outlet temperature (1100C) the maximum combined stress in all cases is less while maintaining the blanket SiC/SiC composite than the assumed conservative limit of 190 and Sic/Pbli interface at a lower temperature MPa, often with significant margin(as a 1000C). The first wall consists of a 4-mm SiCd positive measure of reliability) SiC structural wall on which a 1-mm chemical The activation, decay heat, and waste disposal vapor deposition (CVD) Sic armor layer is analyses performed in support of the ARIF deposited at design are described in Ref [9]. The decay First Wall R685 Fig. 2. Cross-section of ARIES-AT outboard blanket segmPb/17Li at a high outlet temperature (1100 8C) while maintaining the blanket SiCf/SiC composite and SiC/PbLi interface at a lower temperature (/ 1000 8C). The first wall consists of a 4-mm SiCf/ SiC structural wall on which a 1-mm chemical vapor deposition (CVD) SiC armor layer is deposited. 3.1.3. Analysis Detailed analyses of the ARIES-AT blanket were performed and the results are summarized are summarized below [1]: . A tritium-breeding ratio of 1.1 was calculated from 3D neutronics analyses of the power core. . Thermal/hydraulic analyses conservatively as￾suming MHD-laminarized Pb/17Li flow showed that for an average outlet Pb/17Li temperature of 1100 8C, both the maximum SiCf/SiC temperature at the FW and the max￾imum blanket SiC/Pb/17Li interface tempera￾ture at the inner channel wall are maintained at /1000 8C, which satisfy the maximum tem￾perature limits shown in Table 1. The corre￾sponding blanket pressure drop is about 0.25 MPa. . Stress analyses were performed both on the module outer and inner shells indicating that the maximum combined stress in all cases is less than the assumed conservative limit of 190 MPa, often with significant margin (as a positive measure of reliability). . The activation, decay heat, and waste disposal analyses performed in support of the ARIES￾AT design are described in Ref. [9]. The decay Fig. 1. ARIES-AT outboard first wall and blanket segment. Fig. 2. Cross-section of ARIES-AT outboard blanket segment. L. Giancarli et al. / Fusion Engineering and Design 61/62 (2002) 307/318 311