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Further improvements of the water-cooled Pb–17Li blanket
Fusion Engineering and Design 58 – 59 (2001) 523– 527 www.elsevier.com/locate/fusengdes
Further improvements of the water-cooled Pb–17Li blanket M.A. Fu¨tterer a,*, G. Benamati b, I. Ricapito b, L. Giancarli c, G. Le Marois d, A. Li Puma c, Y. Poitevin c, J. Reimann e, J.-F. Salavy c, J. Szczepanski c, G. Vella f, G. Ruvutuso f a
CEA Cadarache, DEN/DER, F-13108 Saint Paul-lez-Durance Cedex, France b ENEA Brasimone, P.O. Box 1, I-40023 Camugnano (BO), Italy c CEA Saclay, DEN/DM2S, F-91191 Gif-sur-Y6ette Cedex, France d CEA Grenoble, DTA-CEREM/DEM, 17, rue des Martyrs, F-38054 Grenoble, France e Forschungszentrum Karlsruhe, IKET, P.O. Box 3640, D-76021 Karlsruhe, Germany f Department of Nuclear Engineering, Uni6ersity of Palermo, I-90128 Palermo, Italy
Abstract The water-cooled lithium–lead (WCLL) blanket is based on reduced-activation ferritic– martensitic steel as the structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new approaches are currently under evaluation. This paper outlines several specific modifications, it highlights progress made on various issues and outlines the R&D work which is still required to define an improved reference design for the WCLL concept. © 2001 Elsevier Science B.V. All rights reserved. Keywords: Lithium–lead blanket; Ferritic–martensitic steel; Fusion power plant
1. Introduction The EU water-cooled lithium – lead (WCLL) blanket was originally developed for DEMO [1]. In the framework of preparatory work for a European power plant study an improved version of the blanket (I-WCLL) was presented [2]. The
* Corresponding author. Tel.: + 33-4-42-2540-50; fax: + 33-4-42-2575-95. E-mail address: [email protected] (M.A. Fu¨tterer).
principal difference in layout is that DEMO had a double-null configuration while a commercial power plant is expected to be based on a singlenull scheme. As in DEMO, the blanket is formed by 48 ‘banana shaped’ outboard segments and 32 straight inboard segments. These segments are essentially steel boxes confining a Pb –17Li pool which is slowly circulated for tritium recovery and chemistry adjustment outside the blanket. Stiffener grids ensure the resistance of the segments against accidental pressurization by the coolant and minimize deformation so as to facilitate ex-
0920-3796/01/$ - see front matter © 2001 Elsevier Science B.V. All rights reserved. PII: S 0 9 2 0 - 3 7 9 6 ( 0 1 ) 0 0 4 8 9 - 6
524
M.A. Fu¨tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
change of a faulted module. Poloidal bundles of hairpin shaped double-wall tubes (DWT) remove the heat from the Pb– 17Li pool. The external walls are cooled with an independent circuit. The headers for both circuits enter from the top of the vacuum vessel. A corrugated geometry on the first wall minimizes thermal stress. The wall coolant is collected in the rear of the segment in poloidally oriented headers. Tritium permeation barriers are currently applied to surfaces in contact with Pb– 17Li to minimize the effort for the presumably costly coolant detritiation. A reduced-activation ferritic– martensitic steel (EUROFER-97) was assumed as the structural material with a temperature limit of 550 °C. The use of oxide-dispersion strengthened (ODS) ferritic–martensitic steel, only in the first wall, was postulated to raise this limit to approximately 650 °C. The lifetime of the blanket is probably limited by the behavior of the structural material under irradiation. This blanket is best combined with a watercooled divertor as this enables the use of one single coolant for the entire reactor. Several solutions for such a divertor were proposed which differ in coolant conditions, heat flux limits and types of issues to be resolved. For a given fusion power, the tolerable heat flux on the divertor (7– 9 MW m − 2) determines the size of the reactor and the maximum neutron wall load on the blanket (B 2 MW m − 2). For an I-WCLL blanket with water-cooled divertor, Table 1 shows some of the resulting reactor data representing an economic optimum for a target unit of 1500 MWe.
2. Envisaged design modifications and R&D requirements The I-WCLL blanket has an interesting potential for further design improvements which should be exploited for the recently refined reactor specifications. These design improvements aim at reducing difficult R&D issues, if possible without creating new ones. Emphasis should be put on technically credible and realistic medium-term solutions. Possible improvements, to be fully evaluated in the future, are listed below.
2.1. Adaptation to single-null plasma A commercial power reactor such as that described in Ref. [2] is likely to be based on a single-null configuration. This enables improved blanket coverage, in particular in the top region so as to maximize tritium production, power extraction, and shielding of the magnets. An attractive approach would be a suitable adaptation of what was developed for the ‘convertible blanket’ for ITER [3] in which ‘topboard’ blanket segments were used to fill the gap between inboard and outboard segments on the top. This requires the coolant headers to move to the rear of the blanket.
2.2. Modified cooling system The I-WCLL blanket is currently equipped with poloidally oriented hairpin shaped double-walled cooling tubes to remove the heat from the Pb– Table 1 Specifications for a DEMO and a power reactor equipped with a WCLL blanket and water-cooled divertor (hot divertor: coolant Tmax =320 °C; cold divertor: coolant TmaxB200 °C)
Fusion power (MW) Major plasma radius (m) Divertor heat load (peak) (MW m−2) Total divertor power (MW) Blanket neutron wall load (average/peak) (MW m−2) Blanket surface heat flux (average/peak) (MW m−2) Blanket energy multiplication factor Total blanket power (MW) Total thermal power (blanket+divertor) (MW) Net thermal efficiency (%) Electric power output without divertor (MWe) Electric power output with cold divertor (MWe) Electric power output with hot divertor (MWe)
DEMO, 1989 [1]
Power plant, 1999 [2]
2200 6.3 N/A
4830 9.9 7
491 2.2/2.75
869 1.9/2.25
0.4/0.5
0.34/0.4
1.18
1.15
2027 2518
3676 4545
34.5 699
33 1213
N/A
1213BPB1500
869
1500
M.A. Fu¨ tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
525
a particular design and analysis effort must be invested to master the thermal-hydraulics by suitable tube arrangements.
2.3. Blanket segmentation
Fig. 1. WCLL blanket module with horizontal cooling tubes.
17Li pool. Each of these tubes is longer than 20 m and produced by a HIP process. Despite very positive results in cooling tube testing [4,5], this raises specific tolerance and quality requirements for the tubes and imposes the use of a very large HIP furnace. An interesting alternative could be to use poloidally oriented C-shaped tubes as currently adopted for the present WCLL ITER-FEAT test blanket module [6]. This modification allows to move the coolant headers to the rear and to gain space for tritium breeding. A further evolution could consist in adopting radial– toroidal instead of poloidal cooling, much in the same way as this is done for the segment box cooling. The cooling pipes would be inserted between radial– toroidal steel plates acting as the stiffeners. Fig. 1 shows such an arrangement. This solution has the additional advantages of reducing somewhat the tube length and of reducing the size of the required HIP equipment. The stiffeners must be properly spaced and perforated to provide a meander-shaped pathway for the Pb– 17Li flow, favorable for reducing magneto-hydrodynamic pressure drops and uncertainties related to convective movements. The use of smaller cooling tubes shall be evaluated to ease bending and to provide additional design flexibility. The total amount of steel in this structure does not change significantly so that losses in terms of TBR are not expected. The cost of this modification is that
The 48 I-WCLL outboard blanket segments are approximately 12 m high and 1.4 m wide in the equatorial level with a double curvature and internal stiffener grids. The sheer size and geometrical complexity of these segments raise considerable challenges for fabrication processes the feasibility of which was partly demonstrated by the current R&D program. However, the aspect of tolerances and play between segments and vacuum vessel as well as the mechanical attachment of the blanket was not yet considered in sufficient detail. The maintenance approach for a fusion power plant [7] determines to a large extent its downtime and availability which impact directly the economic viability of a fusion power plant. To ease access and to speed up maintenance and repair, the replacement of small modules by invessel remote-handling is often preferred to the replacement of larger sectors that requires cutting of the vacuum vessel and sometimes even the particularly time consuming heating and cooling of the toroidal field coils. It is, therefore, expected that a segmentation of the blanket into smaller modules (e.g. three modules each for the inboard and outboard blankets plus the topboard blanket) in a similar way as proposed for the TAURO self-cooled blanket [8] may facilitate fabrication of the structure and ease remote handling (cf. Fig. 2). The cost of this advantage is that it increases the number of welds with a tendency to reduce reliability. However, promising techniques to minimize this effect on the cooling tubes were proposed [9] but are subject to reliability tests. Also, some steel is required to account for the additional segment covers. The effect on TBR requires a refined neutronic analysis. The necessity of a divertor blanket and a baffle is to be assessed.
2.4. Adaptation to new power densities The optimum power densities expected for a commercial power plant are slightly lower than
526
M.A. Fu¨ tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
those for DEMO (cf. Table 1). This means that the proportion of cooling water in the first wall can be reduced which is favorable for tritium breeding. Smaller first wall cooling tubes may be envisaged to better minimize thermal stress in the first wall while their pitches should be adapted to the poloidal power density profile so as to homogenize the temperature gains.
2.5. First wall armor A commercial fusion power plant could require armor material on the first wall for which W is a candidate. However, the application of such armor in the form of tiles or coatings was never included in the design and analysis work for blankets despite the fact that it impacts power deposition and tritium production, cooling as well as safety performance, tritium inventory and lifetime due to plasma/first wall interaction. Furthermore, significant work is yet to be done to prove
Fig. 2. Possible segmentation of the WCLL blanket for a single-null reactor.
the feasibility of including the first wall armor in a fabrication process.
2.6. Use of profiled steel structures When increasing the number of blanket segments, it becomes of particular interest to reduce the mass of the structural steel in favor of a high TBR. In numerous industrial applications (aviation, automobile, plastic packing), profiled structures are used to obtain maximum rigidity at minimum material thickness. Such a solution should be investigated in particular for the end caps of the blanket to ensure their resistance against accidental pressurization. Further TBR margins may be obtained by applying improved design rules to the WCLL blanket and by tolerating a certain deformation of the blanket segment in case of pressurization.
2.7. Power con6ersion As a consequence of the assumed physics, the divertor must absorb as much as 18% of the fusion power which economically imposes its use for electricity generation. Two different watercooled divertor types were proposed [10]: 1. one with relatively low coolant temperatures (Tmax : 200 °C), a high heat flux tolerance (9 MW m − 2) and technology close to ITER; 2. the other with high temperature coolant (Tmax = 315 °C), a lower heat flux tolerance (7 MW m − 2) and larger R&D requirements. It is a little problematic to design steam generators for the high temperature divertor to produce live steam at 7 MPa and 285 °C, which are the outlet conditions for the blanket steam generators as well. All steam generators can then be switched in parallel (cycle efficiency : 33%). The gross plant output then becomes 1500 MWe. In such a PWR type steam cycle, feedwater is preheated by steam heated economizers. This recycled power accounts to typically 36% of the steam generator power. In the case of the IWCLL blanket, the steam generator power is 3676 MW (cf. Table 1) so that the required preheating power would amount to 1323 MW. This economizer power can be partly replaced by the
M.A. Fu¨ tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
power from a low-temperature divertor (869 MW). This reduces the required preheater power to be extracted from the turbine to 454 MW so that the electric power output increases significantly although without reaching the performance of the steam producing divertor. Details depend on the divertor operating conditions. Further work in the circuit design should aim at optimizing the integration of the divertor heat in the power conversion cycle and should account for the pulsed operation of a fusion power plant.
2.8. Tritium management Further design and analysis work should be invested to minimize the requirements for the tritium management of the blanket. This concerns in particular the optimization of coolant and breeder detritiation to minimize or even eliminate the need for tritium permeation barriers. A solution to this problem is proposed in [11] and requires in particular a refined safety analysis (definition of tolerable tritium concentration in the cooling water) with feedback to the ancillary system design, segmentation and confinement.
3. Conclusions Several areas were identified for potential design improvements of the water-cooled lithium– lead blanket concept. The estimated advantages were confronted with possible shortcomings and the required R&D work was outlined. The impact of such modifications on the blanket performance should be fully evaluated in the near future to
527
adapt the WCLL blanket to a fusion power plant and to translate this into a suitably modified test blanket module for ITER. References [1] L. Giancarli et al., Development of the EU water-cooled Pb– 17Li blanket, Fusion Eng. Des. 39 – 40 (1998) 639 – 644. [2] I. Cook et al. (Eds.), Preparatory Work in 1999 (Plant Availability) for a Power Plant Conceptual Study, Final Report, FTSC-P(98), p. 41.7. [3] J. Reimann et al., Convertible liquid metal blankets for ITER with Pb– 17Li as breeder material, Fusion Eng. Des. 27 (1995) 353. [4] L. Cachon, Double-wall tube out-of-pile testing, WP A3.3, in: Annual Report of the Association Euratom/ CEA 1999, CEA DSM/DRFC, and further unpublished results. [5] A.J. Magielsen et al., In-pile tritium permeation characteristics of a double walled tube and a TPB coating, Proc. SOFT-21, Madrid, Spain, September 11 – 15, 2000. [6] L. Giancarli et al., Development of the EU water-cooled Pb– 17Li blanket, Fus. Eng. Des. 39 – 40 (1998) 639 –644. [7] J.-P. Friconneau, PPA 3.2 Maintenance of a commercial fusion power station, assessment on availability, CEA Report DTA-CEREM DPSA/STR/LAM/99.115. [8] H. Golfier et al., Performances of the TAURO blanket system associated with a liquid metal cooled divertor, Proc. ISFNT-5, Rome, Italy, September 20 – 24, 1999. [9] E. Rigal et al., Development of FM steels diffusion bonding technologies for blankets manufacturing applications, Proc. ISFNT-5, Rome, Italy, September 19 –24, 1999. [10] L. Giancarli et al., Performance and feasibility of watercooled high heat flux components, CEA Report DRN/ DMT, December 1999. [11] O. Kveton et al., A water-cooled lithium – lead blanket without tritium permeation barriers: feasibility and economical analysis, Proc. SOFT-21, Madrid, Spain, September 11 – 15, 2000.
Further improvements of the water-cooled Pb–17Li blanket M.A. Fu¨tterer a,*, G. Benamati b, I. Ricapito b, L. Giancarli c, G. Le Marois d, A. Li Puma c, Y. Poitevin c, J. Reimann e, J.-F. Salavy c, J. Szczepanski c, G. Vella f, G. Ruvutuso f a
CEA Cadarache, DEN/DER, F-13108 Saint Paul-lez-Durance Cedex, France b ENEA Brasimone, P.O. Box 1, I-40023 Camugnano (BO), Italy c CEA Saclay, DEN/DM2S, F-91191 Gif-sur-Y6ette Cedex, France d CEA Grenoble, DTA-CEREM/DEM, 17, rue des Martyrs, F-38054 Grenoble, France e Forschungszentrum Karlsruhe, IKET, P.O. Box 3640, D-76021 Karlsruhe, Germany f Department of Nuclear Engineering, Uni6ersity of Palermo, I-90128 Palermo, Italy
Abstract The water-cooled lithium–lead (WCLL) blanket is based on reduced-activation ferritic– martensitic steel as the structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new approaches are currently under evaluation. This paper outlines several specific modifications, it highlights progress made on various issues and outlines the R&D work which is still required to define an improved reference design for the WCLL concept. © 2001 Elsevier Science B.V. All rights reserved. Keywords: Lithium–lead blanket; Ferritic–martensitic steel; Fusion power plant
1. Introduction The EU water-cooled lithium – lead (WCLL) blanket was originally developed for DEMO [1]. In the framework of preparatory work for a European power plant study an improved version of the blanket (I-WCLL) was presented [2]. The
* Corresponding author. Tel.: + 33-4-42-2540-50; fax: + 33-4-42-2575-95. E-mail address: [email protected] (M.A. Fu¨tterer).
principal difference in layout is that DEMO had a double-null configuration while a commercial power plant is expected to be based on a singlenull scheme. As in DEMO, the blanket is formed by 48 ‘banana shaped’ outboard segments and 32 straight inboard segments. These segments are essentially steel boxes confining a Pb –17Li pool which is slowly circulated for tritium recovery and chemistry adjustment outside the blanket. Stiffener grids ensure the resistance of the segments against accidental pressurization by the coolant and minimize deformation so as to facilitate ex-
0920-3796/01/$ - see front matter © 2001 Elsevier Science B.V. All rights reserved. PII: S 0 9 2 0 - 3 7 9 6 ( 0 1 ) 0 0 4 8 9 - 6
524
M.A. Fu¨tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
change of a faulted module. Poloidal bundles of hairpin shaped double-wall tubes (DWT) remove the heat from the Pb– 17Li pool. The external walls are cooled with an independent circuit. The headers for both circuits enter from the top of the vacuum vessel. A corrugated geometry on the first wall minimizes thermal stress. The wall coolant is collected in the rear of the segment in poloidally oriented headers. Tritium permeation barriers are currently applied to surfaces in contact with Pb– 17Li to minimize the effort for the presumably costly coolant detritiation. A reduced-activation ferritic– martensitic steel (EUROFER-97) was assumed as the structural material with a temperature limit of 550 °C. The use of oxide-dispersion strengthened (ODS) ferritic–martensitic steel, only in the first wall, was postulated to raise this limit to approximately 650 °C. The lifetime of the blanket is probably limited by the behavior of the structural material under irradiation. This blanket is best combined with a watercooled divertor as this enables the use of one single coolant for the entire reactor. Several solutions for such a divertor were proposed which differ in coolant conditions, heat flux limits and types of issues to be resolved. For a given fusion power, the tolerable heat flux on the divertor (7– 9 MW m − 2) determines the size of the reactor and the maximum neutron wall load on the blanket (B 2 MW m − 2). For an I-WCLL blanket with water-cooled divertor, Table 1 shows some of the resulting reactor data representing an economic optimum for a target unit of 1500 MWe.
2. Envisaged design modifications and R&D requirements The I-WCLL blanket has an interesting potential for further design improvements which should be exploited for the recently refined reactor specifications. These design improvements aim at reducing difficult R&D issues, if possible without creating new ones. Emphasis should be put on technically credible and realistic medium-term solutions. Possible improvements, to be fully evaluated in the future, are listed below.
2.1. Adaptation to single-null plasma A commercial power reactor such as that described in Ref. [2] is likely to be based on a single-null configuration. This enables improved blanket coverage, in particular in the top region so as to maximize tritium production, power extraction, and shielding of the magnets. An attractive approach would be a suitable adaptation of what was developed for the ‘convertible blanket’ for ITER [3] in which ‘topboard’ blanket segments were used to fill the gap between inboard and outboard segments on the top. This requires the coolant headers to move to the rear of the blanket.
2.2. Modified cooling system The I-WCLL blanket is currently equipped with poloidally oriented hairpin shaped double-walled cooling tubes to remove the heat from the Pb– Table 1 Specifications for a DEMO and a power reactor equipped with a WCLL blanket and water-cooled divertor (hot divertor: coolant Tmax =320 °C; cold divertor: coolant TmaxB200 °C)
Fusion power (MW) Major plasma radius (m) Divertor heat load (peak) (MW m−2) Total divertor power (MW) Blanket neutron wall load (average/peak) (MW m−2) Blanket surface heat flux (average/peak) (MW m−2) Blanket energy multiplication factor Total blanket power (MW) Total thermal power (blanket+divertor) (MW) Net thermal efficiency (%) Electric power output without divertor (MWe) Electric power output with cold divertor (MWe) Electric power output with hot divertor (MWe)
DEMO, 1989 [1]
Power plant, 1999 [2]
2200 6.3 N/A
4830 9.9 7
491 2.2/2.75
869 1.9/2.25
0.4/0.5
0.34/0.4
1.18
1.15
2027 2518
3676 4545
34.5 699
33 1213
N/A
1213BPB1500
869
1500
M.A. Fu¨ tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
525
a particular design and analysis effort must be invested to master the thermal-hydraulics by suitable tube arrangements.
2.3. Blanket segmentation
Fig. 1. WCLL blanket module with horizontal cooling tubes.
17Li pool. Each of these tubes is longer than 20 m and produced by a HIP process. Despite very positive results in cooling tube testing [4,5], this raises specific tolerance and quality requirements for the tubes and imposes the use of a very large HIP furnace. An interesting alternative could be to use poloidally oriented C-shaped tubes as currently adopted for the present WCLL ITER-FEAT test blanket module [6]. This modification allows to move the coolant headers to the rear and to gain space for tritium breeding. A further evolution could consist in adopting radial– toroidal instead of poloidal cooling, much in the same way as this is done for the segment box cooling. The cooling pipes would be inserted between radial– toroidal steel plates acting as the stiffeners. Fig. 1 shows such an arrangement. This solution has the additional advantages of reducing somewhat the tube length and of reducing the size of the required HIP equipment. The stiffeners must be properly spaced and perforated to provide a meander-shaped pathway for the Pb– 17Li flow, favorable for reducing magneto-hydrodynamic pressure drops and uncertainties related to convective movements. The use of smaller cooling tubes shall be evaluated to ease bending and to provide additional design flexibility. The total amount of steel in this structure does not change significantly so that losses in terms of TBR are not expected. The cost of this modification is that
The 48 I-WCLL outboard blanket segments are approximately 12 m high and 1.4 m wide in the equatorial level with a double curvature and internal stiffener grids. The sheer size and geometrical complexity of these segments raise considerable challenges for fabrication processes the feasibility of which was partly demonstrated by the current R&D program. However, the aspect of tolerances and play between segments and vacuum vessel as well as the mechanical attachment of the blanket was not yet considered in sufficient detail. The maintenance approach for a fusion power plant [7] determines to a large extent its downtime and availability which impact directly the economic viability of a fusion power plant. To ease access and to speed up maintenance and repair, the replacement of small modules by invessel remote-handling is often preferred to the replacement of larger sectors that requires cutting of the vacuum vessel and sometimes even the particularly time consuming heating and cooling of the toroidal field coils. It is, therefore, expected that a segmentation of the blanket into smaller modules (e.g. three modules each for the inboard and outboard blankets plus the topboard blanket) in a similar way as proposed for the TAURO self-cooled blanket [8] may facilitate fabrication of the structure and ease remote handling (cf. Fig. 2). The cost of this advantage is that it increases the number of welds with a tendency to reduce reliability. However, promising techniques to minimize this effect on the cooling tubes were proposed [9] but are subject to reliability tests. Also, some steel is required to account for the additional segment covers. The effect on TBR requires a refined neutronic analysis. The necessity of a divertor blanket and a baffle is to be assessed.
2.4. Adaptation to new power densities The optimum power densities expected for a commercial power plant are slightly lower than
526
M.A. Fu¨ tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
those for DEMO (cf. Table 1). This means that the proportion of cooling water in the first wall can be reduced which is favorable for tritium breeding. Smaller first wall cooling tubes may be envisaged to better minimize thermal stress in the first wall while their pitches should be adapted to the poloidal power density profile so as to homogenize the temperature gains.
2.5. First wall armor A commercial fusion power plant could require armor material on the first wall for which W is a candidate. However, the application of such armor in the form of tiles or coatings was never included in the design and analysis work for blankets despite the fact that it impacts power deposition and tritium production, cooling as well as safety performance, tritium inventory and lifetime due to plasma/first wall interaction. Furthermore, significant work is yet to be done to prove
Fig. 2. Possible segmentation of the WCLL blanket for a single-null reactor.
the feasibility of including the first wall armor in a fabrication process.
2.6. Use of profiled steel structures When increasing the number of blanket segments, it becomes of particular interest to reduce the mass of the structural steel in favor of a high TBR. In numerous industrial applications (aviation, automobile, plastic packing), profiled structures are used to obtain maximum rigidity at minimum material thickness. Such a solution should be investigated in particular for the end caps of the blanket to ensure their resistance against accidental pressurization. Further TBR margins may be obtained by applying improved design rules to the WCLL blanket and by tolerating a certain deformation of the blanket segment in case of pressurization.
2.7. Power con6ersion As a consequence of the assumed physics, the divertor must absorb as much as 18% of the fusion power which economically imposes its use for electricity generation. Two different watercooled divertor types were proposed [10]: 1. one with relatively low coolant temperatures (Tmax : 200 °C), a high heat flux tolerance (9 MW m − 2) and technology close to ITER; 2. the other with high temperature coolant (Tmax = 315 °C), a lower heat flux tolerance (7 MW m − 2) and larger R&D requirements. It is a little problematic to design steam generators for the high temperature divertor to produce live steam at 7 MPa and 285 °C, which are the outlet conditions for the blanket steam generators as well. All steam generators can then be switched in parallel (cycle efficiency : 33%). The gross plant output then becomes 1500 MWe. In such a PWR type steam cycle, feedwater is preheated by steam heated economizers. This recycled power accounts to typically 36% of the steam generator power. In the case of the IWCLL blanket, the steam generator power is 3676 MW (cf. Table 1) so that the required preheating power would amount to 1323 MW. This economizer power can be partly replaced by the
M.A. Fu¨ tterer et al. / Fusion Engineering and Design 58–59 (2001) 523–527
power from a low-temperature divertor (869 MW). This reduces the required preheater power to be extracted from the turbine to 454 MW so that the electric power output increases significantly although without reaching the performance of the steam producing divertor. Details depend on the divertor operating conditions. Further work in the circuit design should aim at optimizing the integration of the divertor heat in the power conversion cycle and should account for the pulsed operation of a fusion power plant.
2.8. Tritium management Further design and analysis work should be invested to minimize the requirements for the tritium management of the blanket. This concerns in particular the optimization of coolant and breeder detritiation to minimize or even eliminate the need for tritium permeation barriers. A solution to this problem is proposed in [11] and requires in particular a refined safety analysis (definition of tolerable tritium concentration in the cooling water) with feedback to the ancillary system design, segmentation and confinement.
3. Conclusions Several areas were identified for potential design improvements of the water-cooled lithium– lead blanket concept. The estimated advantages were confronted with possible shortcomings and the required R&D work was outlined. The impact of such modifications on the blanket performance should be fully evaluated in the near future to
527
adapt the WCLL blanket to a fusion power plant and to translate this into a suitably modified test blanket module for ITER. References [1] L. Giancarli et al., Development of the EU water-cooled Pb– 17Li blanket, Fusion Eng. Des. 39 – 40 (1998) 639 – 644. [2] I. Cook et al. (Eds.), Preparatory Work in 1999 (Plant Availability) for a Power Plant Conceptual Study, Final Report, FTSC-P(98), p. 41.7. [3] J. Reimann et al., Convertible liquid metal blankets for ITER with Pb– 17Li as breeder material, Fusion Eng. Des. 27 (1995) 353. [4] L. Cachon, Double-wall tube out-of-pile testing, WP A3.3, in: Annual Report of the Association Euratom/ CEA 1999, CEA DSM/DRFC, and further unpublished results. [5] A.J. Magielsen et al., In-pile tritium permeation characteristics of a double walled tube and a TPB coating, Proc. SOFT-21, Madrid, Spain, September 11 – 15, 2000. [6] L. Giancarli et al., Development of the EU water-cooled Pb– 17Li blanket, Fus. Eng. Des. 39 – 40 (1998) 639 –644. [7] J.-P. Friconneau, PPA 3.2 Maintenance of a commercial fusion power station, assessment on availability, CEA Report DTA-CEREM DPSA/STR/LAM/99.115. [8] H. Golfier et al., Performances of the TAURO blanket system associated with a liquid metal cooled divertor, Proc. ISFNT-5, Rome, Italy, September 20 – 24, 1999. [9] E. Rigal et al., Development of FM steels diffusion bonding technologies for blankets manufacturing applications, Proc. ISFNT-5, Rome, Italy, September 19 –24, 1999. [10] L. Giancarli et al., Performance and feasibility of watercooled high heat flux components, CEA Report DRN/ DMT, December 1999. [11] O. Kveton et al., A water-cooled lithium – lead blanket without tritium permeation barriers: feasibility and economical analysis, Proc. SOFT-21, Madrid, Spain, September 11 – 15, 2000.