- Email: [email protected]
The evolution of US helium-cooled blankets
Fusion Engineering and Design 17 (1991) 123-129 North-Holland
123
The evolution of US helium-cooled blankets C.P.C. W o n g a, E.T. C h e n g b a n d K.R. Schultz a General Atomics, PO Box 85608, San Diego, CA 92186-9784, USA b TSI Research, 225 Stevens Avenue, #110, Solana Beach, CA 92075, USA
This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors.
1. Introduction Since the late 1960s, helium has been proposed as the coolant for fusion power reactors [1]. Helium has the distinct beneficial features of being chemically inert, neutron transparent, and capable of operating at high temperature. These features can lead to relatively safe and high performance fusion blanket designs. Heliumcooled blanket concepts were designed for the generation of electricity, fission materials production (i.e., hybrid reactors), and generation of hydrogen (i.e., synfuel reactors). Different structural metallic alloys (e.g., ferritic steel and vanadium alloy) and ceramic-composite materials (e.g., SiC-composite) were used. Utilization of several different solid and liquid tritium breeding materials was considered. All the helium-cooled blanket concepts that were proposed in the past 20 years in the US can be grouped into four generic blanket design configurations. These design configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. In this paper, four blanket designs that can represent the key design features of these four helium-cooled blanket configurations are reviewed and compared. The design configuration that has the potential to be the best for helium-cooled fusion reactor blankets is identified.
2. Design considerations For a fusion reactor blanket design, the basic requirements are adequate tritium breeding and extrac-
tion, high blanket energy multiplication, high thermalhydraulic performance, and simplicity in maintenance and fabrication. Additional desirable features are to complement the shield for the protection of superconducting magnets, to generate minimum induced radioactivity and afterheat, and to require minimum blanket thickness. These requirements and desirable features can only be fulfilled by the proper choice and combination of structural and tritium breeder materials and the suitable selection of configurational design. Helium, being a gas, has relatively low volumetric heat capacity; therefore, when it is used as the working fluid for power reactors, a high gas pressure of 5 to 10 MPa will be needed to maintain a high heat transfer coefficient and low coolant loop pumping power. The trade-offs of these design requirements, desirable features and properties of the helium coolant, are what the designers need to work on in the evolution of a good blanket design.
3. The pressurized module design For the purpose of operating at 5 to 10 MPa gas pressure, the helium coolant of a fusion reactor blanket is either contained in sturdily designed blanket modules or many small diameter tubes. Many of the early US helium-cooled blanket designs used the pressurized module [2-5] design configuration. A typical design selected for comparison is illustrated in Figs. 1 and 2 [5]. This is an integrated first wall and blanket design evolved during the Blanket Comparison and Selection Study [5]. The key features are the coolant routing and the module structural de-
0 9 2 0 - 3 7 9 6 / 9 1 / $ 0 3 . 5 0 © 1991 - Elsevier Science Publishers B.V. A l l rights reserved
124
C.P. C. Wong et a L / The evolution of US hefium-cooled blankets
OUTLETPIPE(i.2 m DIA) ~
\\
'
SHIELD STRONGBACK END PLATE
,NSULATEO OIVIOE. ~L_
~ E L I U M ~ I~) ] ~"~ } ~L / ~ OUTLET~/----d'~I I~J.~'b yJ '/ COOLANT--~I I ~ ~ /.~-'l=l~ ] I ~ F'~ v \ INLET-OUTLET FIRST MANIFOLD
----~
1L~~ L : ~ - ~
("//~u,~T ~:7~//" 9m°'A)
INLETPIPE
I CH~UlIII IIIIIIIllllllN/LINLET DISTRIBUTION
I
.~"~,%~.~ INLETRINGDUCT(1.8rn D I A ) / ~
Fig. 1. Tokamak helium-cooled pressurized blanket modules.
sign. Helium coolant entering from the back of the b l a n k e t at 2 7 5 ° C is directed initially along side inlet channels to the first wall a n d then radially cross flowed t h r o u g h the lithium b r e e d i n g a n d reflector zones, exit-
Jill
JlIIIIIII / o.,,,,,E.
IHIIIIH
IIIIIW/
IIIIIIIIIInll UlIIIIIII4 OLADOEO L,~O ~LATE
.,.000CT (1.9 ,. o~A)
WALL ~;li~SgAg~S kt
,/
i i ~ It II ILI~ FIRST WALL FLOW ./~ Iflll 8 II II IIII II II U gil CHANNEL / t IIIII II II II II H II II II Mmg .... / PRESSURE
I /
-t----~~~
~L
FINNED FIRST WALL (TOKAMAK)
FLOW DISTRIBUTOF
Fig. 2. Pressurized module design.
Table 1 Key parameters of the four selected gas-cooled blanket designs. Blanket Concept
Pressurized Modules (1984)
Pressurized Tubes (1980)
Particulate + Gas (1988)
Nested Shell (1990)
Reactor Application
Tokamak Electricity production HT-9 ferritic
Mirror Fissile fuel production HT-9 ferritic
Tokamak Electricity production SiCcomposite
Tokamak Electricity production SiCcomposite
ki20
LilvPb~3
Li 2O
Li 2ZrO3
None
Thorium
Beryllium
Beryllium
Helium
Helium
Helium
5 275 510
5 250 450
CO 2 + SiC particulate ~' 0.5 250 700
29
16
9
22
39
36
50
49
Structural material Tritium breeder Neutron multiplier Coolant Conditions: Gas Pressure (MPa) T,,1 (°C) Tout (°C) Blanket pressure drop (kPa) Gross thermal efficiency (%) Tritium breeding ratio 0.5 MPa
1.18
1.1
CO2 mixed with 1.5% (by volume) of 5 to 10 micron SiC particulate.
1.1
10 350 650
1.23
125
C.P. C. Wong et aL / The evolution of US hehum-cooled blankets
A variation of this blanket concept can be represented by the synfuel blanket [4] which uses tubular tritium breeder elements. Alternate solid breeder material, LiAIO2, and liquid metal breeders, lithium and lithium-lead, have also been used [5]. The tubular configuration used in the Doublet Demonstration Power reactor design [2], was based on a cylindrical module, 39 cm in diameter, to contain the high pressure helium instead of the lobe configuration illustrated in Fig. 2. Due to the trade-offs of minimizing the number of modules and the volume fraction of the structural material, this pressurized module design is restricted to operate at only about 5 MPa pressure. This lower pressure requires higher pumping power than would be needed for higher pressure operation. Uncertainty about the erosion of the apexes of the lobes when they are exposed to the plasma is another area of concern.
ing at 510°C. This coolant routing scheme has the distinct advantage of optimizing the cooling by guiding the cooler inlet helium to the first wall and then the blanket internals where the Li20 breeder plates are located. The parameters of this blanket are summarized in Table 1. The most difficult task in designing the helium-cooled pressurized module design is the module itself. The module structure has to withstand the high coolant pressure, while not being so massive as to absorb neutrons that are needed to breed the tritium. This requirement leads to the lobe configuration and strongback design as illustrated in Fig. 2. In the lobe design, the coolant pressure is reacted by the hoop stress in the first wall, while the strong-back structure carries the gas load and provides a high-Z material shield. With careful design of the module ends, the module was designed as a stand-alone unit. For the extraction of tritium, a purge flow design was adopted with a separate stream of helium flowing slowly inside the fuel plates to remove the tritium. This design also has the distinct characteristic of locating most of the weld joints at the back of the blanket module.
,3A OUBLE
.:.° \
\
4. Pressurized tube design The second design configuration to contain the high coolant pressure of helium is the pressurized tube des-
~~AS ,~1/.-~'~ ~
~
,~ ~" ~
///////~ //////II/
//////#// / / l ~ / / / / / J
y/~~',,(////
~
~
v
' / / . ~ / / Z ~ ~
\
,
/
_, ~ ~ f
-
~
-
~
UCTS Li Pb LEVEL
CONTROL
CORRUGATED BOX 3 ATM DESIGN PRESSURE
",,GRAPHITE ZONE - OR SiC NONSTRESSEDBLOCKS MAY BE METAL CLAD. OR COATED
~
o.E.O.W/JH....
iiN~'l~i~&C;l:n LITHIUM LEAD, WITH THORIUM ~;'~,:~;~. . . . m.u~n~
IN SUSPENSION. MIXTURE
CIRCULATED SLOWLY
"CONCENTRIC FLOW COOLING TUBES i
I PLASMA
Fig. 3. Schematic of a gas-cooled beryllium blanket for the Tandem Mirror Hybrid Reactor.
C.P.C. Wong et a L / The evolution of US helium-cooled blankets
126
SHIELD
--,RESSU.E
PLENUM CONCE TUBE nESIGN IN~ " ~ HeINIT
COOUNG
TUBE ARRAY
number of tubes and the corresponding large number of welds at the back of the tube plena. This may have an unfavorable impact on the operational availability of the blanket design.
T ]
~'
5. Particulate and gas mixture blanket design
I
-
/
!,
°0
t
Be
ZONE
[ ,
FIRST-WALL
Fig. 4. Pressurized tube design.
ign, which is also called the breeder-out-of-tube (BOT) design. To minimize the length of the pressurized tubes, an innovative double wall reentrant pressurized tube design (Figs. 3 and 4) [6] was invented for the Tandem Mirror Hybrid Reactor study in 1980. This is a hybrid blanket producing fissile fuel at 0.8 atoms per D - T neutrons, in addition to providing tritium breeding at a ratio of 1.12. The 60 cm x 1.4 cm diameter tubes extend towards the plasma from the rear pressure vessel and constitute the majority of the 4% ferritic steel content in the blanket zone, as shown in Fig. 4. The 5 MPa inlet helium coolant at 250°C flows radially from the inlet plenum through the center of the concentric tube toward the first wall. The coolant turns around at the tip of the tube and flows along the annulus of the concentric tube toward the reflector and then flows into the outlet plenum. The outlet coolant temperature is 450°C. The coolant at the tip of the tube would be higher than 250 ° C because of the recuperation effect from the hot outlet coolant. The recuperation effect was kept to an acceptable level by introducing in the pressure tube an annular 1 mm stagnant helium layer for insulation within two 0.25 mm wall thickness tubes. This design has the distinct advantage of the use of a relatively low pressure blanket module which contains the high pressure coolant tubes and the breeding material. The tritium can be extracted from the solid breeder with a low pressure helium purge and from the liquid breeder by slow circulation of the liquid breeder. The key concern of contact resistance for the BOT configuration was resolved by using the liquid LilvPbs,~. For a tokamak reactor, a separately cooled first wall will be needed. At a triangular pitch of - 4 cm between tubes, the key disadvantage of this pressurized tube design is the large
Recognizing that the key drawback of helium coolant is its low volumetric specific heat, a gas-carried solidparticulate design was developed for ITER in 1988 [7] and was later used as a blanket option for the ARIES-I design [8,9]. SiC composite was used as the blanket structural material, reducing activation and allowing a higher coolant outlet temperature. When solid breeder particles are mixed into the gas stream, the mixture will have a much higher volumetric heat capacity than the pure gas. In comparison, for the same heat transfer coefficient,the mixture of particulate and gas will require a lower volume flow rate, pressure drop, and pumping power than a gas-only design. For the ARIES-I blanket design, both helium and COa carrier gasses were considered. To maintain solid-to-gas mass ratio of less than 15 (a rule-of-thumb for uniform particulate gas mixture flow), a gas pressure of 3 MPa is required for helium. Due to higher gas density, a lower gas pressure of 0.5 MPa is required by a CO 2, thus CO 2 was selected as the carrier gas. The ARIES-I CO 2 particulate blanket design parameters are presented in Table 1. As illustrated in Fig. 5, the blanket coolant flows in the poloidal direction through the first wall of the blanket module from the bottom to the top where it makes a 180 o turn and flows back through the blanket
1 cm LI20 PLATES2.5 mm GAPS
VA.A OM
CRUSH.~
II
--
~....~,...-~__~_/~'",~'~ ~
I
I
I I 1I",.
~
. ....... /
~
5 mm
LI20 _TIKES. . . . . . . ~ V,~PORDEPOSITED SIC COAT
Fig. 5. ARIES-I particulate and ('()~ mixture blanket module design.
127
C.P.C. Wong et al. / The evolution of US hefium-cooled blankets
tritium breeder zone. This flow pattern reduces the possibility of the particle settling within the blanket internals. Potential problems associated with this concept are particle and wall erosion and uniform circulation of particles. Experimental data on the specific particulate and structural material are needed to verify the concept. Because of these uncertainties, a pure helium coolant was selected for the ARIES-I reference design, as described next.
6. Nested shell design
With the lessons learned from the pressurized module, pressurized tube, and particulate gas mixture design studies, a relatively simple nested shell design (Figs. 6 and 7) was invented in 1990 [8] and selected for the ARIES-I [8] reference design. The ARIES-I design uses SiC composite as the structural material. The 10 MPa helium coolant is contained in 0.5 to 0.8 cm diameter channels in U-shaped shells that form the blanket structure. The sphere-pac Li2ZrO 3 solid tritium breeder mixed with neutron multiplier Be metal pellets is placed
Fig. 7. ARIES-I reference nested shell blanket module assembly schematic.
// ENOCOVERP
L
A
T
E
~
f
X i BLANKET
FLAREDSHELLBOTTOM TO FIT PLENUM6LOCKCHANNELS
I~11
P~~OLOIOALPLENA
~COLLECTION PLENUM
SHELL
BREEDER
between the nested shells. This nested shell design has the simplest configuration among the designs discussed in this paper. The reliability of the many individual tube joints is improved by forming the tubes into panels, as illustrated in Fig. 7. There is a total of 17 nested shells to form the blanket. The key advantages of the nested shell concept are its design simplicity and its flexibility in application. This shell configuration can be used for different combinations of metallic and ceramic-composite structural materials and for different solid and liquid metal breeder materials. The nested shell configuration also has the advantage of having the possibility of incorporating a smooth first wall surface facing the plasma. No obvious unresolved critical issue has been identified. For a metallic structure design, this blanket configuration offers much promise as a driver and test module blanket for ITER. A summary of the design is presented in Table 2.
SHELLENO
7. Conclusions
Fig. 6. ARIES-I reference nested shell blanket poloidal module.
The development of helium-cooled blanket concepts in the US in the past 20 years was reviewed and four
128
C.P.C Wong et al. / The evolution of US helium-cooled blankets
Table 2 Summary of the ARIES-I nested shell blanket designs. Solid breeder Multiplier Coolant Configuration Structural Coolant
Li2ZrO 3 and Be sphere-pac mixture of 1.0 and 0.1 mm pellets. Be sphere-pac. Be recycling technology will need to be developed. 10 MPa helium. Layers of solid breeder and Be sphere-pac mixture, coolant channels, Be and SiC reflectors and SiC plenum. 17 nested U-shaped shells forming the poloidal modules. Poloidal flow in the plena, distributed into radial in, toroidal and radial out flow. cooling the first wall and the blanket. Breeder Sphere-pac materials filling up the space between breeder zone shells. Fabrication Preformed U-shaped shells to be fitted into the grooves of the reflector/plenum assembly, one layer after another to form the poloidal module, with internal supports as needed. The outer shell is the first wall. Structural First wall and blanket structural analyses completed. Peak total stress is 72 MPa, well below the 120 MPa design limit for irradiated SiC-composite. analysis Tritium purge Cusps between circular coolant channels form the purge channels, helium purge gas pressure flow design at about flow design 0.4 MPa. Tritium breeding ratio = 1.23 Neutronics Blanket energy multiplication = 1.30 Thermal-hydraulics Coolant channels embedded in the U-shaped shells. Pressure drops and peak material temperatures are acceptable. Total first wall and blanket loop pumping power is 19 MW. Coolant inlet and outlet temperatures are 350 o and 650°C, respectively. Gross power cycle efficiency 49%. Blanket tritium Low, 1 gram in the solid breeder. By heating the blanket to 610°C once a year, the tritium inventory in the Be tritium recoil implantation can be controlled to 0.6 kg. inventory
design configurations were identified. They are the pressurized m o d u l e design, the pressurized tube design, the particulate-gas mixture design, a n d the nested shell design. As indicated in Table 1, all the identified b l a n k e t designs can fulfill the r e q u i r e m e n t s of a fusion reactor blanket, including tritium breeding, a n d can have reasonable thermal-hydraulic properties. T h e pressurized m o d u l e design has the disadvantages of design complexity, the difficulty of designing the b l a n k e t structurally, a n d the u n c e r t a i n t y of designing the lobe configuration facing the plasma. T h e pressurized tube design has the potential disadvantage of having too m a n y tubes a n d welded joints, which can-adversely affect the operational reliability of the design a n d is not well-suited to a high heat flux first wall surface. The particulate-gas mixture design has the potential d i s a d v a n t a g e of particle a n d wall erosion, an area of m u c h u n c e r t a i n t y due to the lack of experimental data. The nested shell design has n o a p p a r e n t unresolved problems. A m o n g the four configurations that were reviewed, it has the simplest design configuration, ease of fabrication, a n d great flexibility in the selection of structural material a n d tritium b r e e d i n g material. T h e nested shell design should be considered as one of the configurations for the I T E R driver b l a n k e t a n d test m o d u l e designs.
Acknowledgements This work was s p o n s o r e d by the U n i t e d States Dep a r t m e n t of Energy u n d e r contract n u m b e r D E - A C 0 3 89ER52153.
References [1] G.R. Hopkins and G. Melese-d'Hospital, Direct helium cooling for a fusion reactor, Proc. of the Conf. on Nuclear Fusion Reactors, Culham, England (1969). pp. 522-535. [2] Fusion Engineering Staff, Doublet demonstration fusion power reactor study, General Atomics Report. GA-A14742 (July 1978). [3] Project Staff, Blanket, shield and power conversion system for a small field-reversed mirror fusion reactor, General Atomics Report, GA-A15533 (July 1979), property of Electric Power Research Institute. [4] R.W. Werner and M.A. Hoffman, The gas-cooled, Li2 ° moderator/breeder canister blanket for fusion-synfuels, Nuclear Technology/Fusion 4 (1983) pp. 1067. [5] C.P.C. Wong, R.F. Bourque and E.T. Cheng et al.. Helium-cooled blanket designs, Fusion Technology 8 (1985) pp. 114.
C.P.C Wong et al. / The evolution of US helium-cooled blankets
[6] E.T. Cheng, R.L. Creedon and C. Maya et al., A fissionsupressed gas-cooled blanket for the TMHR, proc. 9th symposium on engineering problems of fusion research, Chicago, 1981, Vol. II, pp. 1831. [7] C.P.C. Wong, E.T. Cheng and R.L. Creedon et al., A Li-particulate blanket concept for ITER, Fusion Technology 15 (1989) pp. 871. [8] C.P.C. Wong, E.T. Cheng and B. McQuillan et al.,
129
ARIES-I SiC compoiste low activation blanket design, proc. 9th topical meeting on the technology of fusion energy, Oak Brook, 1990, to be published in Fusion Technology. [9] C.P.C. Wong and M.Z. Hasan, Thermal-hydraulic design of a solid particulate fusion reactor blanket, Transaction of the 10th International Conference on SMiRT, Vot. N, (June 1989) p. 19.
123
The evolution of US helium-cooled blankets C.P.C. W o n g a, E.T. C h e n g b a n d K.R. Schultz a General Atomics, PO Box 85608, San Diego, CA 92186-9784, USA b TSI Research, 225 Stevens Avenue, #110, Solana Beach, CA 92075, USA
This paper reviews and compares four helium-cooled fusion reactor blanket designs. These designs represent generic configurations of using helium to cool fusion reactor blankets that were studied over the past 20 years in the United States of America (US). These configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. Among these four designs, the nested shell design, which was invented for the ARIES study, is the simplest in configuration and has the least number of critical issues. Both metallic and ceramic-composite structural materials can be used for this design. It is believed that the nested shell design can be the most suitable blanket configuration for helium-cooled fusion power and experimental reactors.
1. Introduction Since the late 1960s, helium has been proposed as the coolant for fusion power reactors [1]. Helium has the distinct beneficial features of being chemically inert, neutron transparent, and capable of operating at high temperature. These features can lead to relatively safe and high performance fusion blanket designs. Heliumcooled blanket concepts were designed for the generation of electricity, fission materials production (i.e., hybrid reactors), and generation of hydrogen (i.e., synfuel reactors). Different structural metallic alloys (e.g., ferritic steel and vanadium alloy) and ceramic-composite materials (e.g., SiC-composite) were used. Utilization of several different solid and liquid tritium breeding materials was considered. All the helium-cooled blanket concepts that were proposed in the past 20 years in the US can be grouped into four generic blanket design configurations. These design configurations are the pressurized module design, the pressurized tube design, the solid particulate and gas mixture design, and the nested shell design. In this paper, four blanket designs that can represent the key design features of these four helium-cooled blanket configurations are reviewed and compared. The design configuration that has the potential to be the best for helium-cooled fusion reactor blankets is identified.
2. Design considerations For a fusion reactor blanket design, the basic requirements are adequate tritium breeding and extrac-
tion, high blanket energy multiplication, high thermalhydraulic performance, and simplicity in maintenance and fabrication. Additional desirable features are to complement the shield for the protection of superconducting magnets, to generate minimum induced radioactivity and afterheat, and to require minimum blanket thickness. These requirements and desirable features can only be fulfilled by the proper choice and combination of structural and tritium breeder materials and the suitable selection of configurational design. Helium, being a gas, has relatively low volumetric heat capacity; therefore, when it is used as the working fluid for power reactors, a high gas pressure of 5 to 10 MPa will be needed to maintain a high heat transfer coefficient and low coolant loop pumping power. The trade-offs of these design requirements, desirable features and properties of the helium coolant, are what the designers need to work on in the evolution of a good blanket design.
3. The pressurized module design For the purpose of operating at 5 to 10 MPa gas pressure, the helium coolant of a fusion reactor blanket is either contained in sturdily designed blanket modules or many small diameter tubes. Many of the early US helium-cooled blanket designs used the pressurized module [2-5] design configuration. A typical design selected for comparison is illustrated in Figs. 1 and 2 [5]. This is an integrated first wall and blanket design evolved during the Blanket Comparison and Selection Study [5]. The key features are the coolant routing and the module structural de-
0 9 2 0 - 3 7 9 6 / 9 1 / $ 0 3 . 5 0 © 1991 - Elsevier Science Publishers B.V. A l l rights reserved
124
C.P. C. Wong et a L / The evolution of US hefium-cooled blankets
OUTLETPIPE(i.2 m DIA) ~
\\
'
SHIELD STRONGBACK END PLATE
,NSULATEO OIVIOE. ~L_
~ E L I U M ~ I~) ] ~"~ } ~L / ~ OUTLET~/----d'~I I~J.~'b yJ '/ COOLANT--~I I ~ ~ /.~-'l=l~ ] I ~ F'~ v \ INLET-OUTLET FIRST MANIFOLD
----~
1L~~ L : ~ - ~
("//~u,~T ~:7~//" 9m°'A)
INLETPIPE
I CH~UlIII IIIIIIIllllllN/LINLET DISTRIBUTION
I
.~"~,%~.~ INLETRINGDUCT(1.8rn D I A ) / ~
Fig. 1. Tokamak helium-cooled pressurized blanket modules.
sign. Helium coolant entering from the back of the b l a n k e t at 2 7 5 ° C is directed initially along side inlet channels to the first wall a n d then radially cross flowed t h r o u g h the lithium b r e e d i n g a n d reflector zones, exit-
Jill
JlIIIIIII / o.,,,,,E.
IHIIIIH
IIIIIW/
IIIIIIIIIInll UlIIIIIII4 OLADOEO L,~O ~LATE
.,.000CT (1.9 ,. o~A)
WALL ~;li~SgAg~S kt
,/
i i ~ It II ILI~ FIRST WALL FLOW ./~ Iflll 8 II II IIII II II U gil CHANNEL / t IIIII II II II II H II II II Mmg .... / PRESSURE
I /
-t----~~~
~L
FINNED FIRST WALL (TOKAMAK)
FLOW DISTRIBUTOF
Fig. 2. Pressurized module design.
Table 1 Key parameters of the four selected gas-cooled blanket designs. Blanket Concept
Pressurized Modules (1984)
Pressurized Tubes (1980)
Particulate + Gas (1988)
Nested Shell (1990)
Reactor Application
Tokamak Electricity production HT-9 ferritic
Mirror Fissile fuel production HT-9 ferritic
Tokamak Electricity production SiCcomposite
Tokamak Electricity production SiCcomposite
ki20
LilvPb~3
Li 2O
Li 2ZrO3
None
Thorium
Beryllium
Beryllium
Helium
Helium
Helium
5 275 510
5 250 450
CO 2 + SiC particulate ~' 0.5 250 700
29
16
9
22
39
36
50
49
Structural material Tritium breeder Neutron multiplier Coolant Conditions: Gas Pressure (MPa) T,,1 (°C) Tout (°C) Blanket pressure drop (kPa) Gross thermal efficiency (%) Tritium breeding ratio 0.5 MPa
1.18
1.1
CO2 mixed with 1.5% (by volume) of 5 to 10 micron SiC particulate.
1.1
10 350 650
1.23
125
C.P. C. Wong et aL / The evolution of US hehum-cooled blankets
A variation of this blanket concept can be represented by the synfuel blanket [4] which uses tubular tritium breeder elements. Alternate solid breeder material, LiAIO2, and liquid metal breeders, lithium and lithium-lead, have also been used [5]. The tubular configuration used in the Doublet Demonstration Power reactor design [2], was based on a cylindrical module, 39 cm in diameter, to contain the high pressure helium instead of the lobe configuration illustrated in Fig. 2. Due to the trade-offs of minimizing the number of modules and the volume fraction of the structural material, this pressurized module design is restricted to operate at only about 5 MPa pressure. This lower pressure requires higher pumping power than would be needed for higher pressure operation. Uncertainty about the erosion of the apexes of the lobes when they are exposed to the plasma is another area of concern.
ing at 510°C. This coolant routing scheme has the distinct advantage of optimizing the cooling by guiding the cooler inlet helium to the first wall and then the blanket internals where the Li20 breeder plates are located. The parameters of this blanket are summarized in Table 1. The most difficult task in designing the helium-cooled pressurized module design is the module itself. The module structure has to withstand the high coolant pressure, while not being so massive as to absorb neutrons that are needed to breed the tritium. This requirement leads to the lobe configuration and strongback design as illustrated in Fig. 2. In the lobe design, the coolant pressure is reacted by the hoop stress in the first wall, while the strong-back structure carries the gas load and provides a high-Z material shield. With careful design of the module ends, the module was designed as a stand-alone unit. For the extraction of tritium, a purge flow design was adopted with a separate stream of helium flowing slowly inside the fuel plates to remove the tritium. This design also has the distinct characteristic of locating most of the weld joints at the back of the blanket module.
,3A OUBLE
.:.° \
\
4. Pressurized tube design The second design configuration to contain the high coolant pressure of helium is the pressurized tube des-
~~AS ,~1/.-~'~ ~
~
,~ ~" ~
///////~ //////II/
//////#// / / l ~ / / / / / J
y/~~',,(////
~
~
v
' / / . ~ / / Z ~ ~
\
,
/
_, ~ ~ f
-
~
-
~
UCTS Li Pb LEVEL
CONTROL
CORRUGATED BOX 3 ATM DESIGN PRESSURE
",,GRAPHITE ZONE - OR SiC NONSTRESSEDBLOCKS MAY BE METAL CLAD. OR COATED
~
o.E.O.W/JH....
iiN~'l~i~&C;l:n LITHIUM LEAD, WITH THORIUM ~;'~,:~;~. . . . m.u~n~
IN SUSPENSION. MIXTURE
CIRCULATED SLOWLY
"CONCENTRIC FLOW COOLING TUBES i
I PLASMA
Fig. 3. Schematic of a gas-cooled beryllium blanket for the Tandem Mirror Hybrid Reactor.
C.P.C. Wong et a L / The evolution of US helium-cooled blankets
126
SHIELD
--,RESSU.E
PLENUM CONCE TUBE nESIGN IN~ " ~ HeINIT
COOUNG
TUBE ARRAY
number of tubes and the corresponding large number of welds at the back of the tube plena. This may have an unfavorable impact on the operational availability of the blanket design.
T ]
~'
5. Particulate and gas mixture blanket design
I
-
/
!,
°0
t
Be
ZONE
[ ,
FIRST-WALL
Fig. 4. Pressurized tube design.
ign, which is also called the breeder-out-of-tube (BOT) design. To minimize the length of the pressurized tubes, an innovative double wall reentrant pressurized tube design (Figs. 3 and 4) [6] was invented for the Tandem Mirror Hybrid Reactor study in 1980. This is a hybrid blanket producing fissile fuel at 0.8 atoms per D - T neutrons, in addition to providing tritium breeding at a ratio of 1.12. The 60 cm x 1.4 cm diameter tubes extend towards the plasma from the rear pressure vessel and constitute the majority of the 4% ferritic steel content in the blanket zone, as shown in Fig. 4. The 5 MPa inlet helium coolant at 250°C flows radially from the inlet plenum through the center of the concentric tube toward the first wall. The coolant turns around at the tip of the tube and flows along the annulus of the concentric tube toward the reflector and then flows into the outlet plenum. The outlet coolant temperature is 450°C. The coolant at the tip of the tube would be higher than 250 ° C because of the recuperation effect from the hot outlet coolant. The recuperation effect was kept to an acceptable level by introducing in the pressure tube an annular 1 mm stagnant helium layer for insulation within two 0.25 mm wall thickness tubes. This design has the distinct advantage of the use of a relatively low pressure blanket module which contains the high pressure coolant tubes and the breeding material. The tritium can be extracted from the solid breeder with a low pressure helium purge and from the liquid breeder by slow circulation of the liquid breeder. The key concern of contact resistance for the BOT configuration was resolved by using the liquid LilvPbs,~. For a tokamak reactor, a separately cooled first wall will be needed. At a triangular pitch of - 4 cm between tubes, the key disadvantage of this pressurized tube design is the large
Recognizing that the key drawback of helium coolant is its low volumetric specific heat, a gas-carried solidparticulate design was developed for ITER in 1988 [7] and was later used as a blanket option for the ARIES-I design [8,9]. SiC composite was used as the blanket structural material, reducing activation and allowing a higher coolant outlet temperature. When solid breeder particles are mixed into the gas stream, the mixture will have a much higher volumetric heat capacity than the pure gas. In comparison, for the same heat transfer coefficient,the mixture of particulate and gas will require a lower volume flow rate, pressure drop, and pumping power than a gas-only design. For the ARIES-I blanket design, both helium and COa carrier gasses were considered. To maintain solid-to-gas mass ratio of less than 15 (a rule-of-thumb for uniform particulate gas mixture flow), a gas pressure of 3 MPa is required for helium. Due to higher gas density, a lower gas pressure of 0.5 MPa is required by a CO 2, thus CO 2 was selected as the carrier gas. The ARIES-I CO 2 particulate blanket design parameters are presented in Table 1. As illustrated in Fig. 5, the blanket coolant flows in the poloidal direction through the first wall of the blanket module from the bottom to the top where it makes a 180 o turn and flows back through the blanket
1 cm LI20 PLATES2.5 mm GAPS
VA.A OM
CRUSH.~
II
--
~....~,...-~__~_/~'",~'~ ~
I
I
I I 1I",.
~
. ....... /
~
5 mm
LI20 _TIKES. . . . . . . ~ V,~PORDEPOSITED SIC COAT
Fig. 5. ARIES-I particulate and ('()~ mixture blanket module design.
127
C.P.C. Wong et al. / The evolution of US hefium-cooled blankets
tritium breeder zone. This flow pattern reduces the possibility of the particle settling within the blanket internals. Potential problems associated with this concept are particle and wall erosion and uniform circulation of particles. Experimental data on the specific particulate and structural material are needed to verify the concept. Because of these uncertainties, a pure helium coolant was selected for the ARIES-I reference design, as described next.
6. Nested shell design
With the lessons learned from the pressurized module, pressurized tube, and particulate gas mixture design studies, a relatively simple nested shell design (Figs. 6 and 7) was invented in 1990 [8] and selected for the ARIES-I [8] reference design. The ARIES-I design uses SiC composite as the structural material. The 10 MPa helium coolant is contained in 0.5 to 0.8 cm diameter channels in U-shaped shells that form the blanket structure. The sphere-pac Li2ZrO 3 solid tritium breeder mixed with neutron multiplier Be metal pellets is placed
Fig. 7. ARIES-I reference nested shell blanket module assembly schematic.
// ENOCOVERP
L
A
T
E
~
f
X i BLANKET
FLAREDSHELLBOTTOM TO FIT PLENUM6LOCKCHANNELS
I~11
P~~OLOIOALPLENA
~COLLECTION PLENUM
SHELL
BREEDER
between the nested shells. This nested shell design has the simplest configuration among the designs discussed in this paper. The reliability of the many individual tube joints is improved by forming the tubes into panels, as illustrated in Fig. 7. There is a total of 17 nested shells to form the blanket. The key advantages of the nested shell concept are its design simplicity and its flexibility in application. This shell configuration can be used for different combinations of metallic and ceramic-composite structural materials and for different solid and liquid metal breeder materials. The nested shell configuration also has the advantage of having the possibility of incorporating a smooth first wall surface facing the plasma. No obvious unresolved critical issue has been identified. For a metallic structure design, this blanket configuration offers much promise as a driver and test module blanket for ITER. A summary of the design is presented in Table 2.
SHELLENO
7. Conclusions
Fig. 6. ARIES-I reference nested shell blanket poloidal module.
The development of helium-cooled blanket concepts in the US in the past 20 years was reviewed and four
128
C.P.C Wong et al. / The evolution of US helium-cooled blankets
Table 2 Summary of the ARIES-I nested shell blanket designs. Solid breeder Multiplier Coolant Configuration Structural Coolant
Li2ZrO 3 and Be sphere-pac mixture of 1.0 and 0.1 mm pellets. Be sphere-pac. Be recycling technology will need to be developed. 10 MPa helium. Layers of solid breeder and Be sphere-pac mixture, coolant channels, Be and SiC reflectors and SiC plenum. 17 nested U-shaped shells forming the poloidal modules. Poloidal flow in the plena, distributed into radial in, toroidal and radial out flow. cooling the first wall and the blanket. Breeder Sphere-pac materials filling up the space between breeder zone shells. Fabrication Preformed U-shaped shells to be fitted into the grooves of the reflector/plenum assembly, one layer after another to form the poloidal module, with internal supports as needed. The outer shell is the first wall. Structural First wall and blanket structural analyses completed. Peak total stress is 72 MPa, well below the 120 MPa design limit for irradiated SiC-composite. analysis Tritium purge Cusps between circular coolant channels form the purge channels, helium purge gas pressure flow design at about flow design 0.4 MPa. Tritium breeding ratio = 1.23 Neutronics Blanket energy multiplication = 1.30 Thermal-hydraulics Coolant channels embedded in the U-shaped shells. Pressure drops and peak material temperatures are acceptable. Total first wall and blanket loop pumping power is 19 MW. Coolant inlet and outlet temperatures are 350 o and 650°C, respectively. Gross power cycle efficiency 49%. Blanket tritium Low, 1 gram in the solid breeder. By heating the blanket to 610°C once a year, the tritium inventory in the Be tritium recoil implantation can be controlled to 0.6 kg. inventory
design configurations were identified. They are the pressurized m o d u l e design, the pressurized tube design, the particulate-gas mixture design, a n d the nested shell design. As indicated in Table 1, all the identified b l a n k e t designs can fulfill the r e q u i r e m e n t s of a fusion reactor blanket, including tritium breeding, a n d can have reasonable thermal-hydraulic properties. T h e pressurized m o d u l e design has the disadvantages of design complexity, the difficulty of designing the b l a n k e t structurally, a n d the u n c e r t a i n t y of designing the lobe configuration facing the plasma. T h e pressurized tube design has the potential disadvantage of having too m a n y tubes a n d welded joints, which can-adversely affect the operational reliability of the design a n d is not well-suited to a high heat flux first wall surface. The particulate-gas mixture design has the potential d i s a d v a n t a g e of particle a n d wall erosion, an area of m u c h u n c e r t a i n t y due to the lack of experimental data. The nested shell design has n o a p p a r e n t unresolved problems. A m o n g the four configurations that were reviewed, it has the simplest design configuration, ease of fabrication, a n d great flexibility in the selection of structural material a n d tritium b r e e d i n g material. T h e nested shell design should be considered as one of the configurations for the I T E R driver b l a n k e t a n d test m o d u l e designs.
Acknowledgements This work was s p o n s o r e d by the U n i t e d States Dep a r t m e n t of Energy u n d e r contract n u m b e r D E - A C 0 3 89ER52153.
References [1] G.R. Hopkins and G. Melese-d'Hospital, Direct helium cooling for a fusion reactor, Proc. of the Conf. on Nuclear Fusion Reactors, Culham, England (1969). pp. 522-535. [2] Fusion Engineering Staff, Doublet demonstration fusion power reactor study, General Atomics Report. GA-A14742 (July 1978). [3] Project Staff, Blanket, shield and power conversion system for a small field-reversed mirror fusion reactor, General Atomics Report, GA-A15533 (July 1979), property of Electric Power Research Institute. [4] R.W. Werner and M.A. Hoffman, The gas-cooled, Li2 ° moderator/breeder canister blanket for fusion-synfuels, Nuclear Technology/Fusion 4 (1983) pp. 1067. [5] C.P.C. Wong, R.F. Bourque and E.T. Cheng et al.. Helium-cooled blanket designs, Fusion Technology 8 (1985) pp. 114.
C.P.C Wong et al. / The evolution of US helium-cooled blankets
[6] E.T. Cheng, R.L. Creedon and C. Maya et al., A fissionsupressed gas-cooled blanket for the TMHR, proc. 9th symposium on engineering problems of fusion research, Chicago, 1981, Vol. II, pp. 1831. [7] C.P.C. Wong, E.T. Cheng and R.L. Creedon et al., A Li-particulate blanket concept for ITER, Fusion Technology 15 (1989) pp. 871. [8] C.P.C. Wong, E.T. Cheng and B. McQuillan et al.,
129
ARIES-I SiC compoiste low activation blanket design, proc. 9th topical meeting on the technology of fusion energy, Oak Brook, 1990, to be published in Fusion Technology. [9] C.P.C. Wong and M.Z. Hasan, Thermal-hydraulic design of a solid particulate fusion reactor blanket, Transaction of the 10th International Conference on SMiRT, Vot. N, (June 1989) p. 19.