Chapter 13. Reactor Design Studies g Utility criteria for choosing - PowerPoint PPT Presentation

Economy of Scale Issues 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 Investment risk reluctance to invest 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 dolan swip 2009 31

Many Large Power Stations Exist 8 hydroelectric plants 8 hydroelectric plants > 5 GWe > 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? dolan swip 2009 32

Grid Perturbation Avoidance 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 Siting large fusion power plants. dolan swip 2009 33

Fusion could be competititve if 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 . dolan swip 2009 34

Fusion-Fission Hybrids Fusion Fission Hybrids dolan swip 2009 35

Need for Breeding Fissile Fuel Natural Uranium = 0.72% 235 U the rest is 238 U Present reactors use mostly 235 U Breeding 239 Pu: 238 U + neutron  239 Np  239 Pu 2.3 days Estimated World Resources 235 U: 3.88*10 4 tonnes = 2.5x10 21 J = 95 TW-a 238 U: 5.43*10 6 tonnes = 3.5x10 23 J = 1300 TW-a 232 Th: (2.57*10 6 tonnes) = 1.7x10 23 J = 6300 TW-a dolan swip 2009 36

Hybrids better than Liquid Metal Fast Breeder Reactors (LMFBRs) Breeder Reactors (LMFBRs) No fissile fuel required for startup C iti Critical mass avoided l id d 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 dolan swip 2009 37

Main Parameters Blanket energy gain M (energy/neutron) / 14-MeV Example: If n + 6 Li  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 dolan swip 2009 38

Neutronics issues Power density ~ 100 W/cm 3 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 dolan swip 2009 39

Breeding 232 Th  233 U dolan swip 2009 40

Estimates of T, F, M with U 3 Si Blanket dolan swip 2009 41

Ratio of fission to fusion power  = capture to fission ratio  = capture-to-fission ratio C = fission reactor conversion ratio 233 U C = 0.85,  = 0.1  P fis /P fus = 69 239 Pu C = 0.6,  = 0.3  P fis /P fus = 23 dolan swip 2009 42

Tokamak Hybrid Blanket Radiation shield Radiation shield Cooled by Low-pressure water LiFBeF 2 + SS + He LiFBeF 2 SS He dolan swip 2009 43

Catalyzed DD Hybrids dolan swip 2009 44

Advantages of Catalyzed DD Hybrid 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 on-line reprocessing lower inventories of fissile fuel and radioactivity l i t i f fi il f l d di ti it less afterheat lower reprocessing costs Disadvantages: corrosion, low fusion power density dolan swip 2009 45

European Power Plant Designs European Power Plant Designs dolan swip 2009 46

Power Plant Conceptual Study (PPCS). 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 construction time  5 years dolan swip 2009 47

Four Models (A & B) similar to ITER 8 H<1.2, n/n GR <1.2, β N <3.5 A first stability region first stability region. 6 6 B B C Z(m) D 4 (C & D) high β and H ITER 2 2 strong plasma shaping, R(m) high bootstrap current 0 0 5 10 15 fraction fraction -2 -2 divertor protection. -4 -6 -6 -8 (f (from Ian Cook, UKAEA) I C k UKAEA) dolan swip 2009 48

Four Models Parameter Model A Model B Model C Model D Unit Size (GW e ) 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 Bootstrap Fraction 0.45 0 45 0 43 0.43 0.63 0 63 0 76 0.76 P add (MW) 246 270 112 71 Recirculating power fraction g p 0.28 0.27 0.13 0.11 Divertor Peak load (MW/m -2 ) 15 10 10 5 Av. neutron wall load 2.2 2.0 2.2 2.4 (from Ian Cook, UKAEA) dolan swip 2009 49

Materials of Four Models Mode Divertor Blanket structure Breeder/coolant l structure A A RAFM RAFM PbLi + water PbLi water Cu/water or Cu/water or RAFM/water B W + RAFM/He RAFM Li 4 SiO 4 + Be + He C W + RAFM/He RAFM (ODS) PbLI + SiC + He D SiC + PbLi SiCi PbLi All use tungsten alloy divertor armor. From D. Maisonnier 2004 dolan swip 2009 50

Other key parameters Other key parameters Parameter Model A Model B Model C Model D F Fusion i 5.0 3.6 3.4 2.5 power (GW) 20 20 13 5 13.5 30 30 35 35 Q Q Recirculating p power 0.28 0.27 0.13 0.11 fraction Wall load 2.2 2.0 2.2 2.4 (MW/m 2) ( Divertor 15 10 5 peak load 10 (MW/m 2 ) ( ) (from Ian Cook, UKAEA) dolan swip 2009 51

Divertor Lifetime is Important (from Ian Cook, UKAEA) dolan swip 2009 52

Costs: internal and external Costs: internal and external Costs: internal and external Costs: internal and external • Contributions to the cost of electricity: Contributions to the cost of electricity: • Internal costs: constructing fuelling • Internal costs: constructing, fuelling, operating, maintaining, and disposing of power plants of, power plants. • External costs: environmental damage, adverse health impacts. (from Ian Cook, UKAEA) dolan swip 2009 53

Internal costs: scaling Internal costs: scaling • Cost of electricity is 1.5 well represented by the scaling opposite. it 1 PPCS) • The figure shows coe (P systems code systems code calculations for 0.5 Models A to D, against the scaling against the scaling. 0 • Shows that PPCS 0 0.5 1 1.5 Models are good Models are good coe(scaling) coe(scaling) representatives of a 0.6   1 1 1  much wider class of coe   A  0.5 0.4 0.4 0.3   A  η η P P β β N N possible designs. possible designs th e N (from Ian Cook, UKAEA) dolan swip 2009 54

Conventional cost items make the l largest contribution to the externalities t t ib ti t th t liti 1.80E-01 1.60E-01 1.40E-01 o/kW h Model 4 1.20E-01 1.00E-01 Model 5 m Euro 8 00E 02 8.00E-02 Model 6 6.00E-02 4.00E-02 2.00E-02 0.00E+00 g M anufacturing al al al f t - O ccupationa O ccupationa O ccupationa Transport o Local effec dispersion C of routine of m aterials 14 and H-3 m aterials releases exposure accidents accidents building G lobal M (f (from Ian Cook, UKAEA) I C k UKAEA) dolan swip 2009 55

Achievements of PPCS Economic viability of fusion power with plasma f f 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/m 2 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 . Work needed on divertors and maintenance procedures dolan swip 2009 56

UKAEA Conclusions UKAEA Conclusions UKAEA Conclusions UKAEA Conclusions • Economically acceptable fusion power y plants, with major safety and environmental advantages, are accessible by a “fast-track” development of fusion, through ITER without major materials advances. • There is potential for a more advanced • There is potential for a more advanced second generation of power plants. (f (from Ian Cook, UKAEA) I C k UKAEA) dolan swip 2009 57

Japanese Power Plant Designs Japanese Power Plant Designs dolan swip 2009 58

Heliotron Coils l = 2, m = 10 lik like LHD Experiment LHD E i t dolan swip 2009 59

Confinement Scaling Laws Confinement Scaling Laws HELICAL TOKAMAK 10 10 STELL TOK  ATF ASDEX EXP CHS AUG E 1 1 FFHR CMOD   EXP COMPASS HELE E E D3D tau_exp HSR JET LHD 0.1 0.1 JFT2M MHR JT60U SPPS PBXM W7-A 0.01 PDX 0.01 W7-AS TCV TCV 0.001 0.001 .001 .01 .1 1 10 .001 .01 .1 1 10   ELMY              0 . 83 0 . 50 0 . 10 * *           95 95 0 0 . . 71 71 0 0 . . 16 16 0 0 . . 04 04 * * * * ISS ISS E B E B     1 . 93 0 . 63 0 . 41 0 . 08 0 . 23 0 . 97 0 . 0365 ELMY     R P n B I 95 2 . 21 0 . 65 0 . 59 0 . 51 0 . 80 0 . 40 0 . 08 ISS a R P n B E e 2 / 3 E e dolan swip 2009 60

h h First Wall First Wall Blanket Bl Bl Blanket k t k t Shield Shield SC Coil SC Coil Plasma Plasma w w R m R m B max B max R R R p R p B B B t B t R o R o B 0 B 0 dolan swip 2009 61

FFHR2m2 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. dolan swip 2009 62

FFHR2m2 Parameters R 16.0 m a 2.80 m Blanket thickness 1.1 m Ferritic steel and FLIBE B o in plasma 4.43 T B coil 13.0 T P fus 3.0 GWth Neutron wall load 1.3 MW/m 2 Heating power 100 MW 1.9x10 20 m -3 n e n/n sugo 1.5 Z eff 1.35 H ISS95 1.76 T i 16.1 keV  4.1 % dolan swip 2009 63

FFHR2m2 Blanket dolan swip 2009 64

FLIBE Blanket q” ~ 0.1 MW/m 2  T(surface) ~ 1600 K  little tritium retention Be 2 C pebble bed 1 MW/m 2  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 dolan swip 2009 65

Long blanket lifetime is valuable dolan swip 2009 66

Heliotron Reactor Model Heliotron Reactor Model h h h First Wall First Wall Blanket Blanket Shield Shield SC Coil SC Coil Plasma Plasma w w R m R m m m B max B B max B  R p R p B t B t R R R o R o B 0 B 0 dolan swip 2009 67

Small Plasma-Coil Distance is Needed 9.5 9.5 /kWh 9 8 5 8.5 E, Yen RAF-Flibe line 8 Base Case Base Case CO 7.5 7 0.8 1 1.2 1.4 1.6 1.8 Plasma-coil distance  Plasma coil distance  dolan swip 2009 68

High beta is desirable 20 18 R p , m 16 14 12 10 COE, Yen/kWh 8 6 6 0 1 2 3 4 5 6 beta % beta, % dolan swip 2009 69

High Neutron Wall Load is Desirable g 20 18 18 beta = 1% 16 kWh 14 14 OE, Yen/k 12 beta = 2% 10 10 CO beta=3-7% 8 6 4 0 2 4 6 Neutron wall load, MW/m 2 dolan swip 2009 70

High B max is good 20 18 16 R p , m 14 12 COE, Yen/kWh 10 8 6 6 8 10 12 14 16 18 B max , T dolan swip 2009 71

COE vs. Blanket Lifetime 11 10 10 en/kWh 9 COE, Ye 8 C 7 6 0 5 10 15 20 25 30 35 40 Blanket Lifetime, years y dolan swip 2009 72 A. Sagara

Comparison with ITER Original Reduced Heliotron ITER ITER ITER ITER Total coil 20 12 13 mass, kt C Capital cost, i l 10 10 5 8 8 G$ Fusion Fusion 0 5 0.5 0 5 0.5 5 5 power, GW dolan swip 2009 73

COE in Japan p Fission Fission 50 mill/kWh 50 mill/kWh Coal 54 Natural gas Natural gas 58 58 Oil 101 Pumped hydro storage 112 Heliotron base case 72 beta 5%, NLHD, 3GWe 49 dolan swip 2009 74

Conclusions of NIFS Study Neutron wall load ~ 4 MW/m 2 desirable. S Strong economy of scale: High P e  competitive COE f l Hi h P  i i COE Higher   smaller a  lower   ignition difficult Higher   smaller a p  lower  E  ignition difficult Engineering uncertainties Engineering uncertainties coil reliability divertor divertor maintenance dolan swip 2009 75

United States Power Plant Designs United States Power Plant Designs dolan swip 2009 76

Advanced Reactor Innovative Engineering Study (ARIES) 6 75m/1 5m 10MA  = 6.75m/1.5m, 10MA,  = 1990 1990 ARIES-I ARIES I 8.1 8 1 Uses achieved physics Uses achieved physics. First stability First-stability regime P recirc =20% 1.9%, B coil =21 T, SiC+Li 2 ZrO 3 +He, 650 C, 49%,  n =2.5MW/m 2 2 nd stability regime. 7.5m/2.5m, 30Ma,  = 1991 ARIES- D- 3 He fueled 8.6 III tokamak T i =55 keV 24%, B coil =14 T,  n =0.08MW/m 2 P recirc =24% 2 nd stability 1992 1992 ARIES- ARIES- 2 stability II, IV 1993 PULSAR Pulsed tokamak 14m/1 6m  = 5 % 14m/1.6m,  = 5 %, 1994 1994 SPPS SPPS Stellarator Stellarator 7 5 7.5 P recirc P =5% 5%, i B coil =15T, 4 field periods, 32 modular coils, V+Li, 46% 5.5m/1.4m, 11MA,  5 5m/1 4m 11MA  = 1996 1996 7.6 7 6 P recirc P =17% 17% ARIES ARIES- Reversed- Reversed i RS shear 15% B coil =16 T, V+Li, 610 C,  n =5MW/m 2 tokamak dolan swip 2009 77

Advanced Reactor Innovative Engineering Study (ARIES) ¢/kWh Remarks Year Name Type Parameters 3.2m/1.6m, 28MA,  = 1999 7.9 P recirc =34% ARIES- Spherical ST torus 50%, P(Cu coil)=329MW, B coil =7.4 T, ferritic+PbLi, 700 C, 45% 5.2m/1.3m, 13MA,  = 9%, 2000 4.7 ARIES- Advanced YBCO high- AT technology B coil = 11.5, SiC+PbLi, 1100 temperature tokamak C, 59%, superconductor, ITER89-Pconfinement multiplier =2.0, P recirc =14% 7.75m/1.7m, 4MA,  = 2008 7.8 ARIES- Compact High-n, low-T, ISS95 CS CS stellarator t ll t 6.4%, T=6.6keV, B coil = % confinement multiplier f 15T,  n =2.6 =2.0, P recirc ~ 18% (max.5.4)MW/m 2 , ferritic +PbLi +He, Brayton 43%, H He pumping P=183MW. i P 183MW dolan swip 2009 78

ARIES-AT dolan swip 2009 79

Removable Sector dolan swip 2009 80

Advanced Tokamak Features YBCO high-temperature superconductor Internal Transport Barrier SiC t SiC structure t High-temperature PbLi blanket  efficiency = 59% g p y low COE  4.7 ¢/kWh competitive with other energy sources competitive with other energy sources. dolan swip 2009 81

ARIES-AT Development Needs J(r), n(r) 1. Plasma profile control Reversed shear  internal transport barrier  long  E , high beta, and high bootstrap current fraction. 2. Power flow control q“ to wall and divertor < safe limit 3. Disruption avoidance < one per year 4. SiC composites large sizes radiation damage resistance 4. SiC composites large sizes, radiation damage resistance 5. Compatibility SiC + flowing Pb-17Li at 1000 C 6 Heat exchangers between Pb 17Li and He at 1100 C 6. Heat exchangers between Pb-17Li and He at 1100 C (Brayton cycle for high efficiency) 7. High-heat-flux materials for divertor (probably W) dolan swip 2009 82

Summary – Power Plant Designs F Fusion will be economically viable if: i ill b i ll i bl 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 . dolan swip 2009 83

Extra Slides dolan swip 2009 84

Helical Reactors LHD experiment Heliotron reactor Modular Heliotron Compact Compact Stellarator dolan swip 2009 85

Algorithms for B o , n av , and COE B o /B max = 0.476 (80S coil /R o ) 0.4 (m/10) 0.82 (1.2/  c ) 0.05 /R ) 0 4 ( /10) 0 82 (1 2/ ) 0 05 B /B 0 476 (80S n av =  B t  t (  o 2 /(4  o kT av ) av ) av = 0.26 P -0.59 n e  iss 0.51 B 0.83 R 0.65 a 2.21  2/3 0.4 H iss =  E /  iss H iss > 1.5 achieved in LHD COE = [C AC + (C O&M + C SCR + C F )(1+y) Y ]/(8760P e f avail ) + C D&D dolan swip 2009 86

COE vs. Central Temperature 18 16 R p , m 14 14 12 10 10 COE, Yen/kWh 8 6 0 10 20 30 40 Central Temperature T Central Temperature T o , keV keV dolan swip 2009 87

Comparison of Plasma Aspect Ratios Comparison of Plasma Aspect Ratios 8.5 8 R p /a p = 8.1 /kWh Base Case Base Case 7 5 7.5 OE, Yen/ 7 R p /a p = 5.7 R /a = 5 7 CO 6.5 6 6 1 1.5 2 2.5 3 P e , GWe P e , GWe dolan swip 2009 88

COE vs. Coil width/depth 8.4 8.2 n/kWh 8 7 8 7.8 COE, Yen 7.6 7.4 7.2 7 0 0 1 1 2 2 3 3 4 4 w/h dolan swip 2009 89

COE vs. Profile Parameters n(x)/ n o = (1-y ed )(1 – x p ) q [d + (1-d)x 2 ] + y ed o = (1-t ed )(1 – x r ) s + t ed T(x)/ T x = r/r p 8.5 q=0.25 8 en/kWh flat q=1 7.5 q=2 q 2 COE, Ye 7 q=4 peaked 6 5 6.5 6 0 0 2 2 4 4 6 6 8 8 flat peaked s dolan swip 2009 90

Effect of Hollow Density Profiles n(x)/ n o = (1-y ed )(1 – x p ) q [d + (1-d)x 2 ] + y ed x p ) q [d + (1 d)x 2 ] + y n(x)/ n = (1 y )(1 o = (1-t ed )(1 – x r ) s + t ed T(x)/ T 1 2 1.2 8 5 8.5 y d = 1 1.0 0.8 0.8 8.4 0.6 0.6 0 4 0.4 0 4 0.4 h , Yen/kWh q=0.5 0.2 8.3 0.0 0 0.2 0.4 0.6 0.8 1 x x 8 2 8.2 COE, q=1 8.1 8 Ref: J. F. Lyon, 0.4 0.5 0.6 0.7 0.8 0.9 1 hollow hollow ARIES study flat d dolan swip 2009 91

Required H iss for Ignition Required H iss for Ignition 3 5 3.5 R p /<a p >=5.7 H iss 3  c = 1.25  =6%  = 1 15 m  1.15 m 2 5 2.5 4% 2 2% 1.5 1% 1% 1 "base case" 0.5 0 0 1 2 3 4 P e , GWe dolan swip 2009 92

dolan swip 2009 93

Blanket-Shield Comparison Units RAF-Flibe V-Li (SPPS) SiC-PbLi (AT) Inboard FW/BL/SH/VS Inboard FW/BL/SH/VS m m 0.95 0.95 1.29 1.29 1.02 1.02 thickness M$/m 2 0.27 Inboard blanket+shield 0.37 0.25 cost M$/m 2 0 27 Outboard Outboard M$/m 0.27 0.37 0 37 0 34 0.34 blanket+shield costs Coolant outlet C 560 610 1100 Temperature Energy conversion % 40 46 59 Efficiency Thinnest; Thickest & most Highest But lowest But lowest expensive; expensive; efficiency; efficiency; efficiency but might be but expensive made thinner. materials dolan swip 2009 94

Fusion Power Island Mass vs. P e 50 45 40 beta=2% kt 35 35 30 3% 25 4% 20 20 15 6% 5% 10 5 0 0 0 1 1 2 2 3 3 4 4 Pe, GWe dolan swip 2009 95

Strong Economy of Scale Available COE, Yen/kWh 13 beta = 2% 12 12 11 H iss = 2 10 1.7 1 7 3% 9 1.5 8 4% 4% 7 7 6 5% 5 5 4 0 0.5 1 1.5 2 2.5 3 3.5 P Pe, GWe GW dolan swip 2009 96

ISS-95 and NLHD-D1 scalings = 0.26 P -0.59 n e  iss 0.51 B 0.83 R 0.65 a 2.21  2/3 0.4  NLHD = 0.269 P -0.59 n e 0.52 B 1.06 R 0.64 a 2.58 H iss =  E /  iss dolan swip 2009 97

Required H iss for Ignition Required H iss for Ignition 3 5 3.5 R p /<a p >=5.7 H iss 3  c = 1.25  =6%  = 1 15 m  1.15 m 2 5 2.5 4% 2 2% 1.5 1% 1% 1 "base case" 0.5 0 0 1 2 3 4 P e , GWe dolan swip 2009 98

Heliotrons & Modular Coil Stellarators Heliotrons & Modular Coil Stellarators Heliotrons Modular coil stellarators Theoretical beta < 5%, Th ti l b t 5% Potential beta > 5%, needs P t ti l b t 5% d 4% achieved experimental verification. Alpha confinement uncertain p Potentially good alpha y g p confinement Plasma aspect ratio restricted Aspect ratio can vary over wide by  c to approximate range 5 5 -- by  c to approximate range 5.5 range. Low ratios may yield range. Low ratios may yield lower COE. 8.5. Natural helical divertor Local divertors, space problem NLHD D1 scaling favorable NLHD-D1 scaling favorable dolan swip 2009 99

Heliotrons & Modular Coil Stellarators Heliotrons Modular coil stellarators Coil winding accuracy uncertain. Coil winding & alignment to be demonstrated by W-7X. Coil failure probably unfeasible Failed coil or module could be to repair. replaced. Alignment should last for the Coils must be re-aligned after lifetime of the plant removal of a module Lifetime blanket might be Periodic replacement of blanket feasible. modules envisioned. Large ports available for first Port size generally smaller, wall replacement. depends on specific design. Elliptical shape cross section Elli i l h i Odd h Odd shaped cross sections, d i permits close proximity of more complex. blanket and shield. dolan swip 2009 100