RH. Jones et al. I Journal of Nuclear Materials 307-311(2002)1057-1072 ransmutation rates, chemical compatibility a which are an anticipation of successful future R&D. The sion, irradiation creep and crack growth, join most significant assumptions are the following ology and thermal and mechanical transient Recent experimental results and mechanistic and mod- (i The Sic/Sic thermal conductivity at 1000C and materials. predictions show promise of new, improved at end-of-life(EOL)conditions is 20 W/m K. This value is considerably higher than that shown by present-day SiC/SiC. In fact, ent available data on existing industrial 3D Sicr/SiC indicate 2. Current design possibilities and needs at 1000C a value of 15 W/m K in the plane and of 9 W/m K through the thickness [1], without tak- Silicon carbide composites (SiCr/SiC), are being ing into account the effect of irradiation which are considered in future fusion power reactors because their expected to decrease by about a factor 3 the out high temperature properties (=1000C), offer the po of-pile value tential of very high energy conversion efficiency (50% or (in The maximum and minimum acceptable tempera- more). SiCr/SiC composite has been proposed as struc- tures are respectively 1100C and 600C; these tural material for the first wall and blanket in several values need to be confirmed under irradiation for conceptual design studies EOL conditions (iii)Compatibility between Pb-17Li and Sicr/Sic is ac- ceptable at 800C; this statement should be valid 2.1. Proposed blanket concepts after irradiation and at Pb-17Li velocity of few m/s and should also be valid for any brazing mate- The most recent proposals are TAURO in the Eu- ropean Union, ARIES-AT in the United States, and rials in contact with Pb-17Li: available dat firm a good compatibility for static Pb-17Li at DREAM in Japan. The first two concepts are Pb-17Li 800oC for 3000 h self-cooled blankets, while Dream is cooled by 10 MPa (iv) Use of preliminary SicrSiC models and design Helium [1]. Both TAURO and Aries-AT blankets are criteria are not yet validated by experiments: mod- essentially formed by a SiC/Sic box with indirectly els and criteria currently used for metals and de- cooled Fw that acts as a container for the pb-17Li fined in industrial design codes (.g, ASME, which has the simultaneous functions of coolant tritium RCC-MR) are not applicable for Sic/sic struc- breeder, neutron multiplier and, finally, tritium carrier tures Because of the relatively low SiC/sic electrical con (v) The electrical conductivity of Sic /Sic is about 500 ductivity, high Pb-17Li velocity is allowed without Q2m this value would allow sufficiently low needing large coolant pressures (<1.5 MPa). TAURO MHD effects for self-cooled Pb-17Li blanket and blanket is characterized by 2m-high single modules correspond to the presently measured out-of-pile which are reinforced by SiCr-SiC stiffeners ARIES-AT data. This result could, however, be jeopardized is characterized by a coaxial Pb-17Li flow, which occurs by Pb-17Li infiltration in the top layer of SiC: in two 8 m-high boxes inserted one into the other. The SiC: this infiltration could dramatically increase DREAM blanket is characterized by smaller modul the wall electrical conductivity which could quickly (0.5 m of height), each divided in three zones: FW reeding zone and shield; neutron multiplier material become unacceptably high. A SiC coating on SiCr/ Sic is probably sufficient to avoid this kind of effect (Be), tritium breeding material (LiO or other lithium (vi) Acceptably low coolant leakage in case of He-cool- ceramics)and shielding material(Sic)are packed in the g(10 MPa of pressure); very low quantity of He module as small size pebbles of I mm-diameter for Be tolerated in the plasma so SiCr/Sic hermeticity and Li2O, and 10 mm for SiC. The He coolant path need to be ensured by a reliable coating which includes a flow through the pebble beds and a porous should have the same irradiation resistance of the partition wall. These blankets allow very high coolant main structures; no experimental results are ye outlet temperatures and therefore a high energy con version efficiency. The maximum coolant outlet tem- (vii) Possibility of manufacturing relevant shapes with perature is 1100C obtained in the Pb-17Li of the appropriate thickness ranging between I and 6 ARIES-AT blanket which lead to a thermal efficiency of mm. Present requirements appear achievable in 58.5% present day industrial composites; however, mate- rial properties in these conditions need to be exper 2.2. Main assumptions for blanket de imentally verified. vili)Existing methods of joining finite components The TAURO. ARIEs-AT and dream designs have with characteristics similar to the base material been performed g optimistic SiCr/Sic properties good results are already availabletransmutation rates, chemical compatibility and corro￾sion, irradiation creep and crack growth, joining tech￾nology and thermal and mechanical transient behavior. Recent experimental results and mechanistic and mod￾eling based predictions show promise of new, improved materials. 2. Current design possibilities and needs Silicon carbide composites (SiCf /SiC), are being considered in future fusion power reactors because their high temperature properties (’1000
C), offer the po￾tential of very high energy conversion efficiency (50% or more). SiCf /SiC composite has been proposed as struc￾tural material for the first wall and blanket in several conceptual design studies. 2.1. Proposedblanket concepts The most recent proposals are TAURO in the Eu￾ropean Union, ARIES-AT in the United States, and DREAM in Japan. The first two concepts are Pb–17Li self-cooled blankets, while DREAM is cooled by 10 MPa Helium [1]. Both TAURO and ARIES-AT blankets are essentially formed by a SiC/SiC box with indirectly￾cooled FW that acts as a container for the Pb–17Li which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. Because of the relatively low SiCf /SiC electrical con￾ductivity, high Pb–17Li velocity is allowed without needing large coolant pressures (<1.5 MPa). TAURO blanket is characterized by 2m-high single modules which are reinforced by SiCf–SiC stiffeners. ARIES-AT is characterized by a coaxial Pb–17Li flow, which occurs in two 8 m-high boxes inserted one into the other. The DREAM blanket is characterized by smaller modules (0.5 m of height), each divided in three zones: FW, breeding zone and shield; neutron multiplier material (Be), tritium breeding material (Li2O or other lithium ceramics) and shielding material (SiC) are packed in the module as small size pebbles of 1 mm-diameter for Be and Li2O, and 10 mm for SiC. The He coolant path includes a flow through the pebble beds and a porous partition wall. These blankets allow very high coolant outlet temperatures and therefore a high energy con￾version efficiency. The maximum coolant outlet tem￾perature is 1100
C obtained in the Pb–17Li of the ARIES-AT blanket which lead to a thermal efficiency of 58.5%. 2.2. Main assumptions for blanket designs The TAURO, ARIES-AT and DREAM designs have been performed assuming optimistic SiCf /SiC properties which are an anticipation of successful future R&D. The most significant assumptions are the following: iiii(i) The SiCf /SiC thermal conductivity at 1000
C and at end-of-life (EOL) conditions is 20 W/m K. This value is considerably higher than that shown by present-day SiCf/SiC. In fact, present available data on existing industrial 3D SiCf/SiC indicate at 1000
C a value of 15 W/m K in the plane and of 9 W/m K through the thickness [1], without tak￾ing into account the effect of irradiation which are expected to decrease by about a factor 3 the out￾of-pile value. iii(ii) The maximum and minimum acceptable tempera￾tures are respectively 1100
C and 600
C; these values need to be confirmed under irradiation for EOL conditions. ii(iii) Compatibility between Pb–17Li and SiCf/SiC is ac￾ceptable at 800
C; this statement should be valid after irradiation and at Pb–17Li velocity of few m/s and should also be valid for any brazing mate￾rials in contact with Pb–17Li; available data con- firm a good compatibility for static Pb–17Li at 800
C for 3000 h. ii(iv) Use of preliminary SiCf /SiC models and design criteria are not yet validated by experiments; mod￾els and criteria currently used for metals and de- fined in industrial design codes (e.g., ASME, RCC-MR) are not applicable for SiCf /SiC struc￾tures. iii(v) The electrical conductivity of SiCf /SiC is about 500 X1 m1; this value would allow sufficiently low MHD effects for self-cooled Pb–17Li blanket and correspond to the presently measured out-of-pile data. This result could, however, be jeopardized by Pb–17Li infiltration in the top layer of SiCf/ SiC; this infiltration could dramatically increase the wall electrical conductivity which could quickly become unacceptably high. A SiC coating on SiCf / SiC is probably sufficient to avoid this kind of effect. ii(vi) Acceptably low coolant leakage in case of He-cool￾ing (10 MPa of pressure); very low quantity of He is tolerated in the plasma so SiCf /SiC hermeticity need to be ensured by a reliable coating which should have the same irradiation resistance of the main structures; no experimental results are yet available to give indication about this requirement. i(vii) Possibility of manufacturing relevant shapes with appropriate thickness ranging between 1 and 6 mm. Present requirements appear achievable in present day industrial composites; however, mate￾rial properties in these conditions need to be exper￾imentally verified. (viii) Existing methods of joining finite components with characteristics similar to the base material; good results are already available. 1058 R.H. Jones et al. / Journal of Nuclear Materials 307–311 (2002) 1057–1072