Fusion Engineering and design ELSEVIER Fusion Engineering and Design 61-62(2002)307-318 Progress in blanket designs using SiCSic composites L. Giancarli,*, H. Golfier, S. Nishio, R. Raffray, C. Wong, R. Yamada CEA-Saclay, DEN/CPT, 91191 Gif-sur-Yuette, france CEA-Saclay, DENIDMTSERMA, 91191 Gif-sur- Yrette, Fran JAERI, Depa f Fusion Plasma Research, Naka Fusion Research Establishment, 801- Mukouyama, Naka-imachi, Naka-gun Ibaraki-kenn 311-0193, Japan 458 EBU-l1, University of California, San Diego, 9500 Gilman Drice, La Jolla, CA 92093-0417, US.A General Atomics. San Diego, CA 92186-9784. USA This paper summarizes the most recent design activities concerning the use of SicSic composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the tauRo blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium-lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and aries-at blankets are essentially formed by a sicsic box acting as a container for the lithium-lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplie and, finally, tritium carrier. The dream blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li2O (or other lithium ceramics) as breeder material and of Sic as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R&d on SiCr/Sic c 2002 Published by elsevier Science B V. Keywords: Blanket designs: SiC/SiC composites; Self-cooled lithium-lead 1. Introduction The low activation and afterheat levels asso- ciated with SiC!Sic after long-term neutron The use of SiCdSic composites as structural irradiation allow a design of FPR nuclear compo material for in-vessel components permits to show nents showing high safety standards and simplifie the potential of D-T fusion power reactors(FPR) maintenance schemes. In addition, SicSiC has in terms of safety and environmental impact excellent chemical stability at high temperature, which minimizes mobilization of radioactive pro- Corresponding author. Tel. +33-1-69-08-21-37: fax: +33 Furthermore, the high temperature properties of 6908-58-61 SiCHSic improve energy handling capabilities, E-mail address: luciano. giancarli(@cea fr (L. Giancarli allowing the use of high temperature coolant 0920-3796/02/S- see front c 2002 Published by Elsevier Science B.V. PI:S0920-3796(02)00213-2
Progress in blanket designs using SiCf/SiC composites L. Giancarli a,, H. Golfier b , S. Nishio c , R. Raffray d , C. Wong e , R. Yamada c a CEA-Saclay, DEN/CPT, 91191 Gif-sur-Yvette, France b CEA-Saclay, DEN/DMT/SERMA, 91191 Gif-sur-Yvette, France c JAERI, Department of Fusion Plasma Research, Naka Fusion Research Establishment, 801-1 Mukouyama, Naka-machi, Naka-gun, Ibaraki-kenn 311-0193, Japan d 458 EBU-II, University of California, San Diego, 9500 Gilman Drive, La Jolla, CA 92093-0417, USA e General Atomics, San Diego, CA 92186-9784, USA Abstract This paper summarizes the most recent design activities concerning the use of SiCf/SiC composite as structural material for fusion power reactor breeding blanket. Several studies have been performed in the past. The most recent proposals are the TAURO blanket concept in the European Union, the ARIES-AT concept in the US, and DREAM concept in Japan. The first two concepts are self-cooled lithium/lead blankets, while DREAM is an helium-cooled beryllium/ceramic blanket. Both TAURO and ARIES-AT blankets are essentially formed by a SiCf/SiC box acting as a container for the lithium/lead which has the simultaneous functions of coolant, tritium breeder, neutron multiplier and, finally, tritium carrier. The DREAM blanket is characterized by small modules using pebble beds of Be as neutron multiplier material, of Li2O (or other lithium ceramics) as breeder material and of SiC as shielding material. The He coolant path includes a flow through the pebble beds and a porous partition wall. For each blanket, this paper describes the main design features and performances, the most recent design improvements, and the proposed manufacturing routes in order to identify specific issues and requirements for the future R&D on SiCf/SiC. # 2002 Published by Elsevier Science B.V. Keywords: Blanket designs; SiCf/SiC composites; Self-cooled lithium/lead 1. Introduction The use of SiCf/SiC composites as structural material for in-vessel components permits to show the potential of D/T fusion power reactors (FPR) in terms of safety and environmental impact. The low activation and afterheat levels associated with SiCf/SiC after long-term neutron irradiation allow a design of FPR nuclear components showing high safety standards and simplified maintenance schemes. In addition, SiCf/SiC has an excellent chemical stability at high temperature, which minimizes mobilization of radioactive products. Furthermore, the high temperature properties of SiCf/SiC improve energy handling capabilities, allowing the use of high temperature coolant Corresponding author. Tel.: /33-1-69-08-21-37; fax: /33- 1-69-08-58-61 E-mail address: luciano.giancarli@cea.fr (L. Giancarli). Fusion Engineering and Design 61/62 (2002) 307/318 www.elsevier.com/locate/fusengdes 0920-3796/02/$ - see front matter # 2002 Published by Elsevier Science B.V. PII: S 0 9 2 0 - 3 7 9 6 ( 0 2 ) 0 0 2 1 3 - 2
L Giancarli et al Fusion Engineering and Design 61-62(2002)307-318 with the potential for high energy conversion neutron multiplier and breeder, respectively, and efficiency(>50%) SiCSiC for joints(e.g. bolts). The difficulty in this Starting from the different strategies which can case is to keep the same low-activation require- be adopted for FPR safety and from the r&d ment for all the other in-vessel components next section, this paper presents an assessment of reactivity, low afterheat materials should be used the most recent proposals of breeding blanket This strategy has been adopted by the TauRo designs with particular focus on fabrication issues. and ARIES-AT blanket designs, which use low In particular, two self-cooled lithium-lead (SCLL) pressure, low reactivity, low afterheat Pb-17Li as blanket designs, ARIES-AT [ and TAURO [2]. coolant, neutron multiplier and breeder. The and one helium-cooled be/ceramic(HCBC)design, difficulty in this case is to fulfill the same require- DREAM 33], will be considered ments for the other in-vessel components which imply for instance the use Pb-17Li(or equivalent) as coolant for divertor and shield 2. Attractiveness and development risks for SiCr Sic structures 2. 2. High plant efficiency The attractiveness of Fpr breeding blankets using SiCSic structures is based on the achiev- Maximum acceptable working temperature of able high safety standards and high plant effi- SiC/Sic under irradiation is about 1000C. de- ciency. These significant advantages of SiC!Sic have been developed with the aim of exploiting this favorable feature for having high compared to other structural materials can be fully coolant outlet temperature and, as a consequence, exploited by Dy making coherent design choices concerning the other materials required in the high overall plant efficiency. Moreover, high blanket temperature coolant gives the potential of an efficient hydrogen production in combination 2.1. High safety standards with the standard electricity production. The three designs considered in this paper have High safety standards can be potentially aid particular attention to this aspect. In part achieved because of the low short term activation cular, He-coolant outlet temperature in DREAM blanket is about 900C leading to a net thermal and decay heat which minimize accidental releases, efficiency greater than 45%. For the Tauro facilitates the accommodation of loss-of-coolant (LOCA)and loss-of-flow (LOFA)events, and blanket the Pb-17Li parameters have been opti- simplifies maintenance procedure mized in order to reach an outlet temperature of In particular, in order to limit to an acceptable about 950C and a corresponding net thermal level the accidental release of activation products efficiency of about 55%. In case of ARIES-At the two different strategies can be envisaged (4), that choice of having an annular Pb-17Li flow allows is. either to minimize the in-vessel overall activa to reach an outlet temperature of about 1100C tion inventory and control the release, or to leading to a net thermal efficiency as high as 58.5% minimize the available energy within the safety vessel and keep the activation products confined In the first case, all materials present within the 2.3. R&d requirements and development risks vessel should have low activation characteristics and for the SiCsic a minimization of the Present-day SiCASic composites are not ade- impurity contents should be pursued. This strategy quate to be used directly as structure of nuclear has been adopted by the dream blanket design, components. A comparison between measured which uses only low-activation materials, such as properties on present-day Sic!SiC and require- high-pressure He as coolant, and Be and Li2O ments are given in Table 1. In fact, there are some
with the potential for high energy conversion efficiency (/50%). Starting from the different strategies which can be adopted for FPR safety and from the R&D needs for SiCf/SiC structures, summarized in the next section, this paper presents an assessment of the most recent proposals of breeding blanket designs with particular focus on fabrication issues. In particular, two self-cooled lithium/lead (SCLL) blanket designs, ARIES-AT [1] and TAURO [2], and one helium-cooled be/ceramic (HCBC) design, DREAM [3], will be considered. 2. Attractiveness and development risks for SiCf/ SiC structures The attractiveness of FPR breeding blankets using SiCf/SiC structures is based on the achievable high safety standards and high plant efficiency. These significant advantages of SiCf/SiC compared to other structural materials can be fully exploited by making coherent design choices concerning the other materials required in the blanket. 2.1. High safety standards High safety standards can be potentially achieved because of the low short term activation and decay heat which minimize accidental releases, facilitates the accommodation of loss-of-coolant (LOCA) and loss-of-flow (LOFA) events, and simplifies maintenance procedure. In particular, in order to limit to an acceptable level the accidental release of activation products, two different strategies can be envisaged [4], that is, either to minimize the in-vessel overall activation inventory and control the release, or to minimize the available energy within the safety vessel and keep the activation products confined. In the first case, all materials present within the vessel should have low activation characteristics and for the SiCf/SiC a minimization of the impurity contents should be pursued. This strategy has been adopted by the DREAM blanket design, which uses only low-activation materials, such as high-pressure He as coolant, and Be and Li2O as neutron multiplier and breeder, respectively, and SiCf/SiC for joints (e.g. bolts). The difficulty in this case is to keep the same low-activation requirement for all the other in-vessel components. In the second case, only low pressure, low reactivity, low afterheat materials should be used. This strategy has been adopted by the TAURO and ARIES-AT blanket designs, which use low pressure, low reactivity, low afterheat Pb/17Li as coolant, neutron multiplier and breeder. The difficulty in this case is to fulfill the same requirements for the other in-vessel components which imply for instance the use Pb/17Li (or equivalent) as coolant for divertor and shield. 2.2. High plant efficiency Maximum acceptable working temperature of SiCf/SiC under irradiation is about 1000 8C. Designs have been developed with the aim of exploiting this favorable feature for having high coolant outlet temperature and, as a consequence, high overall plant efficiency. Moreover, high temperature coolant gives the potential of an efficient hydrogen production in combination with the standard electricity production. The three designs considered in this paper have paid particular attention to this aspect. In particular, He-coolant outlet temperature in DREAM blanket is about 900 8C leading to a net thermal efficiency greater than 45%. For the TAURO blanket the Pb/17Li parameters have been optimized in order to reach an outlet temperature of about 950 8C and a corresponding net thermal efficiency of about 55%. In case of ARIES-AT the choice of having an annular Pb/17Li flow allows to reach an outlet temperature of about 1100 8C leading to a net thermal efficiency as high as 58.5%. 2.3. R&D requirements and development risks Present-day SiCf/SiC composites are not adequate to be used directly as structure of nuclear components. A comparison between measured properties on present-day SiCf/SiC and requirements are given in Table 1. In fact, there are some 308 L. Giancarli et al. / Fusion Engineering and Design 61/62 (2002) 307/318
L Giancarli et al. Fusion Engineering and Design 61-62(2002)307-318 309 Table I Comparison between SiCSic properties assumed in the analysis and typical measured values on present-day industrial composites Key SiC/Sic properties and parameters SCLL blankets(agreed drEAM blanket Typical measured value 3000kg/m3 2500kg/m3 ≈2500kg/m Porosity 00-300GI 200 GPa 0.16-0.18 20 0.18 pansion coefficient 4×10-°FC 4×10-6FC Thermal conductivity in plane(1000 C) ≈20WmK(EOL) 15 and 60 W/m k R 15 WIm K(BOL) (EOL) Thermal conductivity through thickness (1000 C) <20 W/m K ( EOL) 15 and 60 W/m K ≈7.5W/mK(BOL (EOL) Electrical conductivity 500/Q2 m(under irradia- Not applicable 500/@2m(out of irradia Tensile strength 300 MPa 300 MPa 300 MPa Trans-laminar shear strength 200 MPa Inter-laminar shear strength 44 MPa Maximum allowable tensile stress Not used 200 MPa Maximum allowable temperature(swelling basis) 1000 C ≈1100°C Maximum allowable interface temperature with 1000°c( nowing) 800°C( statIc) breeder Minimum allowable temperature( thermal conduc- 600C 600° tivity basis) Cost ≤S400/kg ≈10 times larger Assumed design criteria are slight different for each design. They are given in the appropriate chapters. No validated experimental data are yet available key nfluencing its attractiveness, which can development, testing and validation of accep be identified as 'development risks' and which table joining techniques. Different joining tech define the required r&d program. Most R&D niques can be envisaged: (i) assembling by requirements on SiCSic are common to both He- sewing at textile stage to join the stiffeners to cooled and Pb-l7Li cooled systems [5]. The most the side walls; (ii) sticking and co-infiltration to Important common requirements are join the second wall to the first stiffener;(iii) brazing of finished components to join the improvement of thermal conductivity, espe bottom and the top closure plates and the cially through the thickness, at high tempera different sub modules. A promising brazing under technique using a braze material compatible determination and possible improvement of with SiC, the Brasic@, is currently under devel- maximum working temperature under irradia- 46 tion(swelling, compatibility ); development and validation of appropriate de- Specific r&d items concerning SCll blankets sign criteria (e.g. maximum allowed stresses) are which could ensure reasonable component re liability; determination of the electrical conductivity determination and improvement of the lifetime under irradiation capability of fabrication of components with establishment of the maximum interface tem homogeneous properties and reasonable dimen- perature with Pb-17Li under representative sions, with particular attention to the minimum flowing conditions and irradiation level; in particular verification that no Pb-17Li infiltra
key issues influencing its attractiveness, which can be identified as ‘development risks’ and which define the required R&D program. Most R&D requirements on SiCf/SiC are common to both Hecooled and Pb/17Li cooled systems [5]. The most important common requirements are: . improvement of thermal conductivity, especially through the thickness, at high temperature and under neutron irradiation; . determination and possible improvement of maximum working temperature under irradiation (swelling, compatibility); . development and validation of appropriate design criteria (e.g. maximum allowed stresses) which could ensure reasonable component reliability; . determination and improvement of the lifetime; . capability of fabrication of components with homogeneous properties and reasonable dimensions, with particular attention to the minimum and maximum thickness; . development, testing and validation of acceptable joining techniques. Different joining techniques can be envisaged: (i) assembling by sewing at textile stage to join the stiffeners to the side walls; (ii) sticking and co-infiltration to join the second wall to the first stiffener; (iii) brazing of finished components to join the bottom and the top closure plates and the different sub modules. A promising brazing technique using a braze material compatible with SiC, the Brasic†, is currently under development [4,6]. Specific R&D items concerning SCLL blankets are: . determination of the electrical conductivity under irradiation; . establishment of the maximum interface temperature with Pb/17Li under representative flowing conditions and irradiation level; in particular verification that no Pb/17Li infiltraTable 1 Comparison between SiCf/SiC properties assumed in the analysis and typical measured values on present-day industrial composites Key SiCf/SiC properties and parametersa SCLL blankets (agreed values) DREAM blanket Typical measured value Density :/3000 kg/m3 2500 kg/m3 :/2500 kg/m3 Porosity :/5% :/10% :/10% Young’s modulus 200/300 GPa :/200 GPa :/200 GPa Poisson’s ratio 0.16/0.18 0.20 0.18 Thermal expansion coefficient :/4/106 /8C 3.3/106 /8C 4/106 /8C Thermal conductivity in plane (1000 8C) :/20 W/m K (EOL) 15 and 60 W/m K (EOL) :/15 W/m K (BOL) Thermal conductivity through thickness (1000 8C) :/20 W/m K (EOL) 15 and 60 W/m K (EOL) :/7.5 W/m K (BOL) Electrical conductivity :/500/V m (under irradiation) Not applicable :/500/Vm (out of irradiation) Tensile strength 300 MPa 300 MPa 300 MPa Trans-laminar shear strength / / 200 MPa Inter-laminar shear strength / / 44 MPa Maximum allowable tensile Stress Not useda 200 MPaa Unknowna Maximum allowable temperature (swelling basis) :/1000 8C :/1100 8C :/1000 8C Maximum allowable interface temperature with breeder :/1000 8C (flowing) / :/800 8C (static) Minimum allowable temperature (thermal conductivity basis) :/600 8C :/600 8C :/600 8C Cost 0/$400/kg / :/10 times larger a Assumed design criteria are slight different for each design. They are given in the appropriate chapters. No validated experimental data are yet available. L. Giancarli et al. / Fusion Engineering and Design 61/62 (2002) 307/318 309
310 L Giancarli et al. Fusion Engineering and Design 61-62(2002)307-318 tion through the SiC SiC surface will occur, the technologies and of physics understanding major risk being an increase of the wall and modeling capa on the performance of electrical conductivity advanced tokamak plants [7]. The blanket compatibility of brazing material with Pb-17Li. design was developed to achieve high performance while maintaining attractive safety features, simple Specific r&d items concerning HCBC blankets design geometry, credible maintenance and fabri cation processes, and reasonable design margins as hermeticity to high-pressure Helium; an indication of reliability [1] compatibility with Be and Li2O The Pb-I7Li operating temperature is opti- mized to provide high power cycle efficiency while Most of these issues were addressed in detail in maintaining the SiCrSic temperature under rea- presentations and discussions at the January 2000 sonable limits. The Brayton cycle offers the best International Town Meeting on SiC/SiC Design near-term possibility of power conversion with and Material Issues for Fusion Systems and in a high efficiency and is chosen to maximize the related publication [ 5] potential gain from high temperature operation of the Pb-17Li which after exiting the blanket is routed through a heat exchanger with the cycle He 3. Self-cooled Pb-17Li blankets as secondary fluid [8]. The maximum He cycle temperature is 1050C, resulting in a high cycle The safety strategy for SCll blankets is based efficiency of about 58.5% on the minimization of the energy inventory in the The Sic/sic parameters and properties used in vessel. This strategy, in principle, allows the use of he ARIEs-AT analysis are summarized in Table materials for in-vessel components which are not I. For thermo-mechanical analyses it has been low activation(except for long-term waste man- assumed that the maximum allowed combined agement considerations) and which could be of stress(primary and thermal stresses)is 190 MPa particular interest for developing high erior mance joining techniques or for designing other 3.1.2. Blanket description components(such as divertor or shielding) For waste minimization and cost saving reasons. scli blankets use the eutectic Pb-17Li whose the blanket is subdivided radially into two zones: a eplaceable first zone in the inboard and outboard melting point is 235C. Because of the high and a life of plant second zone in the outboard To coolant temperature, it is probably necessary to have the whole coolant circuit made of ceramics simplify the cooling system and minimize the composites(SiC,SiC or equivalent) and it is then number of coolants the pb-17Li is used to cool required to develop specific Pb-17Li/helium heat the blanket as well as the divertor and hot shield exchanger regions. As illustrated in Fig. I and Fig. 2 for the Among advantages one can also note the outboard region, the blanket design is modular and consists of an assembly of simple annular relatively easy tritium extraction to be performed boxes through which the Pb-17Li flows in two outside the reactor and the use of only two basic materials,SiC/SiC and the liquid Pb-17Li which. poloidal passes. Positioning ribs are attached to at least in theory, should allow to reach good he inner annular wall forming a free-floating liability assembly inside the outer wall. These ribs divide he annular region into a number of channels through which the coolant first flows at high 3.1. ARIES-AT blanket velocity to keep cooled both inner and outer walls The coolant then makes a U-turn and flows very 3. .1. General background slowly as a second pass through the large inner The ARIES-at power plant was evolved channel from which the Pb-17Li exits at high assess and highlight the benefit of advanced temperature. This flow scheme enables operatin
tion through the SiCf/SiC surface will occur, the major risk being an increase of the wall electrical conductivity; . compatibility of brazing material with Pb/17Li. Specific R&D items concerning HCBC blankets are: . hermeticity to high-pressure Helium; . compatibility with Be and Li2O. Most of these issues were addressed in detail in presentations and discussions at the January 2000 International Town Meeting on SiCf/SiC Design and Material Issues for Fusion Systems and in a related publication [5]. 3. Self-cooled Pb/17Li blankets The safety strategy for SCLL blankets is based on the minimization of the energy inventory in the vessel. This strategy, in principle, allows the use of materials for in-vessel components which are not low activation (except for long-term waste management considerations) and which could be of particular interest for developing high-performance joining techniques or for designing other components (such as divertor or shielding). SCLL blankets use the eutectic Pb/17Li whose melting point is 235 8C. Because of the high coolant temperature, it is probably necessary to have the whole coolant circuit made of ceramics composites (SiCf/SiC or equivalent) and it is then required to develop specific Pb/17Li/helium heat exchanger. Among the advantages one can also note the relatively easy tritium extraction to be performed outside the reactor and the use of only two basic materials, SiCf/SiC and the liquid Pb/17Li which, at least in theory, should allow to reach good reliability. 3.1. ARIES-AT blanket 3.1.1. General background The ARIES-AT power plant was evolved to assess and highlight the benefit of advanced technologies and of new physics understanding and modeling capabilities on the performance of advanced tokamak power plants [7]. The blanket design was developed to achieve high performance while maintaining attractive safety features, simple design geometry, credible maintenance and fabrication processes, and reasonable design margins as an indication of reliability [1]. The Pb/17Li operating temperature is optimized to provide high power cycle efficiency while maintaining the SiCf/SiC temperature under reasonable limits. The Brayton cycle offers the best near-term possibility of power conversion with high efficiency and is chosen to maximize the potential gain from high temperature operation of the Pb/17Li which after exiting the blanket is routed through a heat exchanger with the cycle He as secondary fluid [8]. The maximum He cycle temperature is 1050 8C, resulting in a high cycle efficiency of about 58.5%. The SiCf/SiC parameters and properties used in the ARIES-AT analysis are summarized in Table 1. For thermo-mechanical analyses it has been assumed that the maximum allowed combined stress (primary and thermal stresses) is 190 MPa. 3.1.2. Blanket description For waste minimization and cost saving reasons, the blanket is subdivided radially into two zones: a replaceable first zone in the inboard and outboard, and a life of plant second zone in the outboard. To simplify the cooling system and minimize the number of coolants, the Pb/17Li is used to cool the blanket as well as the divertor and hot shield regions. As illustrated in Fig. 1 and Fig. 2 for the outboard region, the blanket design is modular and consists of an assembly of simple annular boxes through which the Pb/17Li flows in two poloidal passes. Positioning ribs are attached to the inner annular wall forming a free-floating assembly inside the outer wall. These ribs divide the annular region into a number of channels through which the coolant first flows at highvelocity to keep cooled both inner and outer walls. The coolant then makes a U-turn and flows very slowly as a second pass through the large inner channel from which the Pb/17Li exits at high temperature. This flow scheme enables operating 310 L. Giancarli et al. / Fusion Engineering and Design 61/62 (2002) 307/318
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 segm
Pb/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 assuming 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 maximum blanket SiC/Pb/17Li interface temperature at the inner channel wall are maintained at /1000 8C, which satisfy the maximum temperature limits shown in Table 1. The corresponding 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 ARIESAT 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