Chapter 13. Reactor Design Studies g Utility criteria for choosing - - PowerPoint PPT Presentation
Chapter 13. Reactor Design Studies g Utility criteria for choosing a power plant Reliability, availability, and maintainability Economics Economy of scale Fusion-fission hybrids y European power plant designs Japanese power plant designs
Utility criteria for choosing a power plant Reliability, availability, and maintainability Economics Economy of scale Fusion-fission hybrids y European power plant designs Japanese power plant designs Japanese power plant designs Chinese power plant designs U it d St t l t d i
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United States power plant designs
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Simplicity C it l t Capital cost Construction time Lifetime – 60 years? Fuel-cycle technology, costs, supply Reliability Availability – fraction of time operational y p Load following Maintainability
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Maintainability
Ri k t t kh ld & fi i Risk to stockholders & financiers Staff – size, skills Market – nearby industry or city Transmission lines Grid stability Resources – Nb, He, … Natural hazards – earthquake, flood, fire, wind Waste Decommissioning Absence of fission products, reprocessing
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p p g
Regulations – promote or hinder? Law suits – intervenors? Safety – avoid engineered safety systems? y g y y Emissions – low? Existing regulations? Emergency planning – evacuation plant? Emergency planning evacuation plant? Worker exposure – doses? Li i b f itt d? Licensing – before money committed?
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Public attitude – favorable or hostile? Environmental issues – Environmental issues restrict plant operations? water temperature rise ? water temperature rise ? protected species or archaeological sites? Emissions toxic metals and chemicals radioactivity? Emissions – toxic metals and chemicals, radioactivity? Radioactive waste – minimization, disposal? P bli h lth i t ? Public concerns – health impacts? Perception – public education
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terminology (“emergency core cooling”)
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Reliability = probability that a component will not fail during a certain time period during a certain time period Needed for economical operation Improves safety Related to component failure rates
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N = number of components t = time interval component failure rate (Number of failures)/Nt = component failure rate = (Number of failures)/Nt Reliability during time T = exp(- T) Example: 2 failures among 192 flashlamps
instead of time units T.) = 2 / (192x20,0000) = 5x10-7 per pulse During 50 000 pulses: During 50,000 pulses: Reliability = exp(-5.2x10-7 x 50,000) = 0.974 or 97.4%.
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Availability A = (hours operating)/(total hours in a year) = uptime / (uptime + downtime) = MTBF / (MTBF + MTTR) / ( ) Mean time Mean time to repair between failures Fission reactors A = 75% 90% Capacity factor = (fractional power) x Availability
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Example: Tokamak components operate 80% of the time, Example: Tokamak components operate 80% of the time, 8 hours/day, 5 days/week, 25 weeks/a. A = (0.8 x 8 x 5 x 25)/(24 x 365) = 0.09 = 9% ITER goal = 0.75 hr x (3000 pulses/a) / (24 x 365) = 26%.
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Mean Time to Repair MTTR = Sum(repair times) / (number malfunctions) Example: Instrument repair times 4 0 5 9 2 25 man hours Example: Instrument repair times 4, 0.5, 9, 2.25 man-hours. MTTR = (4 + 0.5 + 9 + 32.25) / 5 = 3.5 hr/repair Mean Down Time MDT = (analyze +get parts +repair +install +test +startup)
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Preventive maintenance done during operations or normal shutdown time. (oil change on car) Predictive maintenance uses instruments to detect signs and predict failures before they occur. (meas re ibrations from bearings) (measure vibrations from bearings)
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Traditional – technicians, electricians, welders, mechanics, carpenters, … “Balance of Plant’ = steam generators, turbines, generators, condenser, pipes, valves, instruments, controls, ... condenser, pipes, valves, instruments, controls, ... “Hands-on” maintenance dose rate < 10 Sv/hr Remote Maintenance – robotic, remote manipulators. heat, radiation, toxic atmosphere. ITER JET divertor modules: 112 hr/week x 15 weeks
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h d l d years
& maintenance
s c
scheduled component replacement Total capital cost Fuel years construction
h e
($ / kWh) f “fi d h t ”
“current dollar mode” E = 0 05 fcr = 0 15
fcr = “fixed charge rate” = cost of capital, depreciation, insurance, taxes. E = “escalation rate” = rate of inflation
current dollar mode E = 0.05, fcr = 0.15 “constant dollar mode” E = 0, fcr = 0.10
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Cc = total capital cost ≈ 1.35 Cdc (FIDC+FEDC)
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Purpose: Compare tokamaks, heliotrons, modular stellarators. Input - Rp/ap, , T
plasma parameters and power balance Adjusts Rp to match desired Pe Masses & unit costs ($/kg) capital cost COE COE variations with plasma and engineering parameters.
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Rp/<ap> 5.7
p p
2 3 % Pe 1.94 GWe Pe 1.94 GWe Bmax 13 T Jmax 30 MA/m2 40 % 40 % Mpol/Mhel 0.40 Msup/Mcoil 0.50
p
T
keV favail 0.75
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% direct M$ % direct
12.7 0.3 21 Structures & site facilities 450 3 11 1
450.3 11.1
2798.2 68.7 22.1 fusion reactor equipment 2090.2 51.3 22 1 1 FW/bl k t/ fl t 259 4 6 4 22.1.1 FW/blanket/reflector 259.4 6.4 22.1.2 shield 254.3 6.2 22.1.3 magnets 956.7 23.5 22.2.4 current drive & heating 168.8 4.1 22.1.5 primary structure & support 169.3 4.2 22 1 6 vacuum systems 198 9 4 9 22.1.6 vacuum systems 198.9 4.9 22.1.7 power supply 67.6 1.7 22.1.8 impurity control & divertor 15.2 0.4 22 1 10 ECRH breakdown system 4 9 0 1
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22.1.10 ECRH breakdown system 4.9 0.1
M$ % 22.2 main heat transport systems 523 12.8 22.3 auxiliary cooling system 6.8 0.2 22.3 auxiliary cooling system 6.8 0.2 22.4 radioactive waste management 12.1 0.3 22.5 fuel handling and storage 108.2 2.7 22 6 other reactor plant eqt 11 0 3 22.6 other reactor plant eqt. 11 0.3 22.7 instrumentation and control 46.8 1.1 22BOP (sum of 22.2 to 22.7) 708 17.4
509.6 12.5
199.6 4.9
100.6 2.5
100.6 2.5
103.4 2.5
4071 100
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Magnet coil ~ 24% of the direct capital cost g p HTSC could reduce this cost. Other fusion reactor components 4 6% each: Other fusion reactor components 4-6% each: Blanket Shield Heating systems Structure Vacuum system y Balance of plant 17%
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COE capital cost 59.751 COE operations 7.875 COEfuel 0 019 COE fuel 0.019 COE replacement 3.387 COE Decon. & decom. 0.612 Total COE 71.6
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Rp 14.4 m 2 1 MW/m2 n 2.1 MW/m M(fusion island) 23 kt Total capital cost 7.9 G$ capital cost / Watt 4.1 $/W COE 72 mill/kWh COE 72 mill/kWh
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intense neutron sources for materials testing. carbon tax on fossil fuel use. If compact high power-density fusion reactors If compact high power density fusion reactors (such as spheromaks) successfully developed fusion fission hybrids built fusion-fission hybrids built large sizes (> 1 Gwe) to achieve economy of scale.
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14
NLHD-D1 Scaling
11 12 13 Wh =2% 3%
w = 1.5 MW/m2
8 9 10 11 E, Yen/kW 3% 4%
2 MW/m2 3 MW/m2
5 6 7 8 COE 5% 4 5 0.5 1 1.5 2 2.5 3 3.5
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Pe, GWe
Lower operating & maintenance costs per kW L i t t li Large equipment cost scaling Geometric effect Pe/(reactor mass) improves with size Lower recirculating power fraction R lt COE COE (P/P ) n 0 4 Result COE = COEo (P/Po)-n n 0.4
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Electricity at one site many plants > 4 GWe Heat rejection at one site many plants > 6 GWth j y p Public acceptance depends on undestanding Investment risk reluctance to invest Investment risk reluctance to invest Market penetration slow, depends on perceive demand Load following generate hydrogen with part of power Grid perturbation site several large plants together Small is beautiful but requires expensive storage Divertor heat load increases with size
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8 hydroelectric plants > 5 GWe 8 hydroelectric plants > 5 GWe Three Gorges Dam (China) = 18.2 GWe ( 2009) 9 nuclear power stations > 4 GWe. New European PWR = 1.6 GWe, single reactor (Limited by control and safety issues) Heliotron Base Case = 1.94 GWe But what about grid perturbation during outage?
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5 large reactors at one site, each g , 60% hydrogen, 40% electricity to grid. Outage of one reactor: Outage of one reactor: 4 reactors, each 50% hydrogen, 50% electricity to grid. Same electrical power to grid.
Ref: J. Sheffield, FST 40 (2001) 1-26 Siting large fusion power plants
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Siting large fusion power plants.
developed for space propulsion intense neutron sources for materials testing. carbon tax on fossil fuel use carbon tax on fossil fuel use. compact high power-density fusion reactors (such as spheromaks) successfully developed (such as spheromaks) successfully developed fusion-fission hybrids built large sizes (> 1 Gwe) to achieve economy of scale.
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Natural Uranium = 0.72% 235U the rest is 238U Present reactors use mostly 235U Breeding 239Pu: 238U + neutron
239Np 239Pu
2.3 days Estimated World Resources
235U: 3.88*104 tonnes = 2.5x1021 J = 95 TW-a 238U: 5.43*106 tonnes = 3.5x1023 J = 1300 TW-a 232Th: (2.57*106 tonnes) = 1.7x1023 J = 6300 TW-a
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No fissile fuel required for startup C iti l id d Critical mass avoided Fuel doubling times months, instead of many years One hybrid supplies many fission reactors Lower power density in uranium fuel Less afterheat safer Accelerates development of pure fusion Accelerates development of pure fusion But Not et de eloped Not yet developed Require tritium-breeding Maintenance more complex
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Blanket energy gain M (energy/neutron) / 14-MeV Example: If n + 6Li 3 8 MeV Example: If n + Li 3.8 MeV then M = (14+3.8)/14 = 1.27 Tritium breeding ratio T tritons bred / incident neutron Fissile fuel breeding ratio F fissile atoms bred / incident neutron
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Power density ~ 100 W/cm3 Blanket always subcritical Neutron multiplier Remove Np and Pu Remove Np and Pu Fuel management (fissile inventory, residence time) Afterheat Afterheat Radiation damage Coolant that does not slow down neutrons Low-pressure coolant minimum structure
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= capture to fission ratio = capture-to-fission ratio C = fission reactor conversion ratio
233U C = 0.85, = 0.1
Pfis/Pfus = 69
239Pu C = 0.6, = 0.3
Pfis/Pfus = 23
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Radiation shield Radiation shield Cooled by Low-pressure water LiFBeF2 + SS + He
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LiFBeF2 SS He
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Tritium breeding not needed more neutrons for fissile fuel breeding can allow more structure structure between first wall, breeder maintenance easier maintenance easier Molten salt
l i t i f fi il f l d di ti it lower inventories of fissile fuel and radioactivity less afterheat lower reprocessing costs Disadvantages: corrosion, low fusion power density
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Safety / Environment no need for emergency evacuation, no active systems for safe shut-down, no active systems for safe shut down, no structure melting following LOCA minimum waste and transport Operation steady state, ~ 1 GWe, base load; availability > 75 % Economics public acceptance Licensing easier than for fission construction time 5 years
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construction time 5 years
(A & B) similar to ITER H<1.2, n/nGR<1.2, βN<3.5 first stability region
6 8
A B
first stability region. (C & D) high β and H
2 4 6
Z(m)
B C D ITER
strong plasma shaping, high bootstrap current fraction
2 5 10 15
R(m)
fraction divertor protection.
(f I C k UKAEA)
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(from Ian Cook, UKAEA)
Parameter Model A Model B Model C Model D Unit Size (GWe) 1.5 1.3 1.4 1.5 Fusion Power (GW) 5.0 3.6 3.4 2.5 Net efficiency 0.31/0.33 0.36 0.41 0.60 Major Radius (m) 9.55 8.6 7.5 6.1 Plasma Current (MA) 30.5 28.0 20.1 14.1 Bootstrap Fraction 0 45 0 43 0 63 0 76 Bootstrap Fraction 0.45 0.43 0.63 0.76 Padd (MW) 246 270 112 71 Recirculating power fraction 0.28 0.27 0.13 0.11 g p Divertor Peak load (MW/m-2) 15 10 10 5
2.2 2.0 2.2 2.4
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(from Ian Cook, UKAEA)
Mode l Divertor structure Blanket structure Breeder/coolant A Cu/water or RAFM PbLi + water A Cu/water or RAFM/water RAFM PbLi water B W + RAFM/He RAFM Li4SiO4 + Be + He C W + RAFM/He RAFM (ODS) PbLI + SiC + He D SiC + PbLi SiCi PbLi
All use tungsten alloy divertor armor.
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From D. Maisonnier 2004
Parameter Model A Model B Model C Model D
F i Fusion power (GW)
5.0 3.6 3.4 2.5
Q
20 13 5 30 35
Q
20 13.5 30 35
Recirculating power
0.28 0.27 0.13 0.11
p fraction Wall load (MW/m2)
2.2 2.0 2.2 2.4
( Divertor peak load (MW/m2)
15 10 10 5
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( ) (from Ian Cook, UKAEA)
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(from Ian Cook, UKAEA)
Contributions to the cost of electricity:
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(from Ian Cook, UKAEA)
well represented by the scaling it
1.5
systems code
1 PPCS)
systems code calculations for Models A to D, against the scaling
0.5 coe (P
against the scaling.
Models are good
0.5 1 1.5 coe(scaling)
Models are good representatives of a much wider class of possible designs
coe(scaling)
0.3 0.4 0.4 0.5 0.6
N β P 1 η 1 A 1 coe
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possible designs.
N e th
N β P η A
(from Ian Cook, UKAEA)
8 00E 02 1.00E-01 1.20E-01 1.40E-01 1.60E-01 1.80E-01
Model 4 Model 5 0.00E+00 2.00E-02 4.00E-02 6.00E-02 8.00E-02 g f al t
al m Euro Model 6 M anufacturing
Transport o m aterials O ccupationa building accidents Local effec
releases G lobal dispersion C 14 and H-3 O ccupationa exposure O ccupationa accidents M
(f I C k UKAEA)
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(from Ian Cook, UKAEA)
f f Economic viability of fusion power with plasma confinement only slightly better than the ITER physics basis. A maintenance concept that could facilititate 75% availability A divertor concept capable of handling 10 MW/m2 A divertor concept capable of handling 10 MW/m Passively safe, so no offsite evacuation plan would be needed Possibility of recycling practically all the materials, with no wastes requiring permanent disposal. q g p p Work needed on divertors and maintenance procedures
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Work needed on divertors and maintenance procedures.
second generation of power plants.
(f I C k UKAEA)
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(from Ian Cook, UKAEA)
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HELICAL TOKAMAK
1 10 ASDEX AUG CMOD COMPASS TOK
1 10 ATF CHS FFHR HELE STELL
EXP E
EXP E
0.01 0.1 D3D JET JFT2M JT60U PBXM PDX TCV
0.01 0.1 tau_exp HSR LHD MHR SPPS W7-A W7-AS
E
0.001 .001 .01 .1 1 10 TCV
0.001 .001 .01 .1 1 10
10 . 50 . 83 .
* *
ELMY
04 . 16 . 71 . 95
* *
ISS
40 . 3 / 2 80 . 51 . 59 . 65 . 21 . 2 95
08 . B n P R a
e ISS E
97 . 23 . 08 . 41 . 63 . 93 . 1
0365 . I B n P R
e ELMY E
B E 04 . 16 . 71 . 95
* *
B ISS E
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First Wall Bl k t
h
First Wall Bl k t
h
SC Coil Blanket Shield
Plasma
w
SC Coil Blanket Shield
Plasma
w
Bmax B R Rm Bmax B R Rm Ro Bt Rp B0 Ro Bt Rp B0
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Varies the pitch of helical coil to reduce Varies the pitch of helical coil to reduce the hoop force on helical coil support structure. Carbon armor tiles containing beryllium soften the neutron energy spectrum, facilitating a local breeding ratio of 1.2.
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R 16.0 m a 2.80 m Blanket thickness 1.1 m Ferritic steel and FLIBE Bo in plasma 4.43 T Bcoil 13.0 T Pfus 3.0 GWth Neutron wall load 1.3 MW/m2 Heating power 100 MW ne 1.9x1020 m-3 n/nsugo 1.5 Zeff 1.35 HISS95 1.76 Ti 16.1 keV 4.1 %
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q” ~ 0.1 MW/m2 T(surface) ~ 1600 K little tritium retention Be2C pebble bed 1 MW/m2 flowing FLIBE coolant Swelling limits C tile lifetime. Manipulator on helical tracks replaces armor tiles remotely Radiation damage limits tile lifetime Rest of RAF blanket lasts 40 years. y
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h
First Wall Blanket Shield
h
First Wall Blanket Shield
h
SC Coil
Plasma
w
SC Coil
Plasma
w
B Rm B Rm R Bmax Bt Rp
m
R Bmax Bt Rp
m
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Ro B0 Ro B0
9.5 8 5 9 9.5 /kWh 8 8.5 E, Yen
Base Case RAF-Flibe line
7 7.5 CO
Base Case
0.8 1 1.2 1.4 1.6 1.8 Plasma-coil distance
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Plasma coil distance
20 16 18 Rp, m 12 14 6 8 10 COE, Yen/kWh 6 1 2 3 4 5 6 beta %
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beta, %
18 20 14 16 18 kWh
beta = 1%
10 12 14 OE, Yen/k
beta = 2%
6 8 10 CO
beta=3-7%
4 2 4 6
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Neutron wall load, MW/m2
20 16 18 Rp, m 12 14 8 10 COE, Yen/kWh 6 6 8 10 12 14 16 18
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Bmax, T
10 11 9 10 en/kWh 8 COE, Ye 6 7 C
5 10 15 20 25 30 35 40
Blanket Lifetime, years
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y
Original ITER Reduced ITER Heliotron ITER ITER Total coil mass, kt 20 12 13 C i l 10 8 Capital cost, G$ 10 5 8 Fusion 0 5 0 5 5 Fusion power, GW 0.5 0.5 5
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Fission 50 mill/kWh Fission 50 mill/kWh Coal 54 Natural gas 58 Natural gas 58 Oil 101 Pumped hydro storage 112 Heliotron base case 72 beta 5%, NLHD, 3GWe 49
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Neutron wall load ~ 4 MW/m2 desirable. S f l Hi h P i i COE Strong economy of scale: High Pe competitive COE Higher smaller a lower ignition difficult Higher smaller ap lower E ignition difficult Engineering uncertainties Engineering uncertainties coil reliability divertor divertor maintenance
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1990 ARIES I First stability 6 75m/1 5m 10MA = 8 1 Uses achieved physics
Advanced Reactor Innovative Engineering Study (ARIES)
1990 ARIES-I First-stability regime 6.75m/1.5m, 10MA, = 1.9%, Bcoil=21 T, SiC+Li2ZrO3+He, 650 C, 49%, n=2.5MW/m2 8.1 Uses achieved physics. Precirc =20% 1991 ARIES- III D-3He fueled tokamak 7.5m/2.5m, 30Ma, = 24%, Bcoil=14 T, n=0.08MW/m2 8.6 2nd stability regime. Ti=55 keV Precirc =24% 1992 ARIES- 2nd stability 1992 ARIES- II, IV 2 stability 1993 PULSAR Pulsed tokamak 1994 SPPS Stellarator 14m/1 6m = 5 % 7 5 P
i
=5% 1994 SPPS Stellarator 14m/1.6m, = 5 %, Bcoil=15T, 4 field periods, 32 modular coils, V+Li, 46% 7.5 Precirc 5%, 1996 ARIES- Reversed- 5 5m/1 4m 11MA = 7 6 P
i
=17% 1996 ARIES RS Reversed shear tokamak 5.5m/1.4m, 11MA, 15% Bcoil=16 T, V+Li, 610 C, n=5MW/m2 7.6 Precirc 17%
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Advanced Reactor Innovative Engineering Study (ARIES)
Year Name Type Parameters ¢/kWh Remarks 1999 ARIES- ST Spherical torus 3.2m/1.6m, 28MA, = 50%, P(Cu coil)=329MW, Bcoil=7.4 T, ferritic+PbLi, 7.9 Precirc =34% 700 C, 45% 2000 ARIES- AT Advanced technology tokamak 5.2m/1.3m, 13MA, = 9%, Bcoil= 11.5, SiC+PbLi, 1100 C, 59%, 4.7 YBCO high- temperature superconductor, ITER89-Pconfinement multiplier =2.0, Precirc =14% 2008 ARIES- CS Compact t ll t 7.75m/1.7m, 4MA, = % 7.8 High-n, low-T, ISS95 f CS stellarator 6.4%, T=6.6keV, Bcoil= 15T, n=2.6 (max.5.4)MW/m2 , ferritic +PbLi +He, Brayton 43%, H i P 183MW confinement multiplier =2.0, Precirc ~ 18% He pumping P=183MW.
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YBCO high-temperature superconductor Internal Transport Barrier SiC t t SiC structure High-temperature PbLi blanket efficiency = 59% g p y low COE 4.7 ¢/kWh competitive with other energy sources competitive with other energy sources.
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J(r), n(r) Reversed shear internal transport barrier long E, high beta, and high bootstrap current fraction.
6 Heat exchangers between Pb 17Li and He at 1100 C
(Brayton cycle for high efficiency)
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F i ill b i ll i bl if Fusion will be economically viable if: neutron source needed for materials testing carbon tax on fossil fuel use compact high power-density fusion reactors (such as spheromaks) successfully developed fusion-fission hybrids built large sizes economy of scale.
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LHD experiment Heliotron reactor Compact Modular Heliotron
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Compact Stellarator
B /B 0 476 (80S /R )0 4 ( /10)0 82 (1 2/ )0 05 Bo/Bmax = 0.476 (80Scoil/Ro)0.4 (m/10)0.82 (1.2/c)0.05 nav = Bt
2/(4okTav) av
t ( o av)
iss = 0.26 P-0.59 ne
0.51 B0.83 R0.65 a2.21 2/3 0.4
Hiss = E/iss Hiss > 1.5 achieved in LHD COE = [CAC + (CO&M + CSCR + CF)(1+y)Y]/(8760Pe favail) + CD&D
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18 14 16 Rp, m 10 12 14 8 10 COE, Yen/kWh 6 10 20 30 40 Central Temperature T keV
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Central Temperature To, keV
8.5 7 5 8 /kWh
Rp/ap = 8.1 Base Case
7 7.5 OE, Yen/
R /a = 5 7 Base Case
6 6.5 CO
Rp/ap = 5.7
6 1 1.5 2 2.5 3 Pe, GWe
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Pe, GWe
8.4 7 8 8 8.2 n/kWh 7.4 7.6 7.8 COE, Yen 7 7.2 1 2 3 4 1 2 3 4 w/h
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n(x)/ no = (1-yed)(1 – xp)q [d + (1-d)x2] + yed T(x)/ T
x = r/rp
8.5 7.5 8 en/kWh
q=0.25 q=1 q 2
flat
6 5 7 COE, Ye
q=2 q=4
peaked
6 6.5 2 4 6 8
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2 4 6 8 s
flat peaked
n(x)/ n = (1 y )(1 xp)q [d + (1 d)x2] + y
8 5
1 2
n(x)/ no = (1-yed)(1 – xp)q [d + (1-d)x2] + yed T(x)/ T
8.4 8.5 h
0 4 0.6 0.8 1.0 1.2 y
d = 1 0.8 0.6 0 4
8 2 8.3 , Yen/kWh q=0.5
0.0 0.2 0.4 0.2 0.4 0.6 0.8 1 x
0.4
8.1 8.2 COE, q=1
x
8 0.4 0.5 0.6 0.7 0.8 0.9 1
Ref: J. F. Lyon, hollow
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d
ARIES study hollow flat
3 5 2 5 3 3.5 Hiss
=6%
Rp/<ap>=5.7 c = 1.25 = 1 15 m
1.5 2 2.5
4% 2% 1%
1.15 m
0.5 1
1% "base case"
1 2 3 4 Pe, GWe
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Units RAF-Flibe V-Li (SPPS) SiC-PbLi (AT) Inboard FW/BL/SH/VS m 0.95 1.29 1.02 Inboard FW/BL/SH/VS thickness m 0.95 1.29 1.02 Inboard blanket+shield cost M$/m2 0.27 0.37 0.25 Outboard M$/m2 0 27 0 37 0 34 Outboard blanket+shield costs M$/m 0.27 0.37 0.34 Coolant outlet Temperature C 560 610 1100 Energy conversion Efficiency % 40 46 59 Thinnest; But lowest Thickest & most expensive; Highest efficiency; But lowest efficiency expensive; but might be made thinner. efficiency; but expensive materials
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50 35 40 45 kt
beta=2%
20 25 30 35
3% 4%
10 15 20
5% 6%
5 1 2 3 4
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1 2 3 4 Pe, GWe
12 13 COE, Yen/kWh
beta = 2%
10 11 12
Hiss = 2 1 7
7 8 9
3% 4%
1.7 1.5
5 6 7
4% 5%
4 5 0.5 1 1.5 2 2.5 3 3.5 P GW
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Pe, GWe
iss = 0.26 P-0.59 ne
0.51 B0.83 R0.65 a2.21 2/3 0.4
NLHD = 0.269 P-0.59 ne
0.52 B1.06 R0.64 a2.58
Hiss = E/iss
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3 5 2 5 3 3.5 Hiss
=6%
Rp/<ap>=5.7 c = 1.25 = 1 15 m
1.5 2 2.5
4% 2% 1%
1.15 m
0.5 1
1% "base case"
1 2 3 4 Pe, GWe
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Heliotrons Modular coil stellarators Th ti l b t 5% P t ti l b t 5% d Theoretical beta < 5%, 4% achieved Potential beta > 5%, needs experimental verification. Alpha confinement uncertain Potentially good alpha p y g p confinement Plasma aspect ratio restricted by c to approximate range 5 5 -- Aspect ratio can vary over wide
by c to approximate range 5.5 8.5.
lower COE. Natural helical divertor Local divertors, space problem NLHD D1 scaling favorable NLHD-D1 scaling favorable
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Heliotrons Modular coil stellarators Coil winding accuracy uncertain. Coil winding & alignment to be demonstrated by W-7X. Coil failure probably unfeasible to repair. Failed coil or module could be replaced. Alignment should last for the lifetime of the plant Coils must be re-aligned after removal of a module Lifetime blanket might be Periodic replacement of blanket feasible. modules envisioned. Large ports available for first wall replacement. Port size generally smaller, depends on specific design. Elli i l h i Odd h d i Elliptical shape cross section permits close proximity of blanket and shield. Odd shaped cross sections, more complex.
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0 degree 45 degree 90 degree plasma shield blanket coil shield QAR (N=2)
dolan swip 2009 101
dolan swip 2009 102
(from D. Maisonnier, 2004
dolan swip 2009 103