Index

‘Note: Page numbers followed by “f” indicate figures and “t” indicate tables.’
A
Active gas handling system (AGHS), 219
Advanced Material Tokamak Experiment (AMTEX) program, 168
Alcator C-Mod
carbon walls, 300–301
compressional-Alfven wave, 313
core turbulence and transport
confinement database, 309
energy transport, 311–312, 312f
multimachine scaling, 309
nonlinear gyrokinetic turbulence models, 309
particle and impurity transport, 311
quantitative analysis, 309
self-generated flows and momentum transport, 310–311, 310f
divertor regimes, 302–304, 303f
D-T fuel cycle, 300
erosion and transport, 301, 302f
high-field operation, 296–297, 296t, 297f
high-field pilot plant concept (ARC), 315–316, 316f
high magnetic field, 295
high-temperature superconductors (HTS), 315–316
H-mode operation, 314
hydrogenic atoms, 301–302
ICRF antenna geometry, 300–301
ICRF wave simulations, 313, 314f
impurity sources, 314–315, 315f
internal hardware, 299, 299f
ion cyclotron range of frequencies (ICRF), 313
L-H transitions, 313
magnets, 298, 298t
material erosion and tritium retention, 300
pedestal and edge barrier regimes
H-modes, 306–308, 307f
I-modes, 308–309, 308f
reactor-scale device, 306
plasma–material interface, 300
turbulence and anomalous transport, boundary plasma, 304–306, 304f–305f
ARIES-CS power plant
components, 592–593, 593f
modular coil set, 592–593, 592f
Axial symmetric divertor experiment (ASDEX) Upgrade, 17–18, 18f
electron cyclotron–current drive (ECCD), 93–94
ITER baseline operation, 111–112, 112f
L–H transition power threshold, 93–94
medium-sized tokamak (MST), 93
MHD modes and disruptions
electron cyclotron–current drive (ECCD), 109
high thermal and mechanical loads, 110–111
massive gas injection (MGI), 109
NTM suppression, 109, 110f
plasma current evolution, 111, 111f
runaway electrons (REs), 111
thermal/current quench, 110–111
pedestal and H-mode physics
core transport, 104, 104f
L–H transition, 100, 101f
linear peeling–ballooning stability, 100–101
pedestal top temperature, 100–101
trans-Greenwald density, 102–104, 102f–103f
plasma–facing components (PFCs), 93–95
plasma wall interaction
divertor retention, 98
edge localized modes (ELMs), 98
gases retention, 96–97
H-mode discharges, 99
ionization stages, 98, 99f
magnetic perturbation coils, 98
neoclassical effects, 98
tungsten plasmaefacing components, 94f, 95–96, 96f
W transport, 99
plasma wall interaction (PWI) project, 93–94
power exhaust, See Power exhaust
q-profile, 114
technical features, 94–95
TORBEAM calculations, 114
type I ELMy H-modes, 113–114, 113f
B
Balance of plant (BOP)
Brayton cycle, 79
energy storage system (EES), 80–81
industrial steam cycle (Rankine) layout, 79, 80f
intermediate heat exchanger (IHX), 81
intermediate heat transport and storage system (IHTS), 80–81
light water reactor technology, 81, 82f
steam turbines, 79, 80f
tokamak reactor, 80–81
Blanket concepts, 61–64, 62t–63t
Bootstrap current density, 13
Breeding blanket, 69–70
Broader Approach (BA) Satellite Tokamak Program, 439
C
Compact toroid (CT) technology, 540–542, 540f
Contextual constraints, magnetic fusion power plants
access, 566
availability, 567–568, 568t
containment, 567
neutral beam-based diagnostics, 568–569
shielding, 567
D
Dual Coolant, 68–69
E
Edge coherent mode (ECM), 415, 416f
Edge localized modes (ELMs), 21, 98
control
electron pressure, 380, 382f
lobe structures, 379–380, 379f
operational space, 378, 379f
parameters, 378
radial profiles, 380, 381f
transient heat loads, 377–378
density perturbation measurements, 372–374, 373f
filamentary structure, 371–372, 372f
fusion reactor, 220–221
heat flux, 374, 374f–375f
high-heat loads, 371–372
high-performance long-pulse operation, 415, 422–425, 422f–425f
pedestal evolution
edge current profiles, 377, 378f
kinetic electron profiles, 375–376, 376f
normalised linear growth rate, 377, 377f
type-I ELM instability, 375–376
spiral pattern, 374
Thomson scattering, 372–375
transients and 3D effects, 41–44, 43f–44f
Edge transport barriers (ETBs), 174
Electron Bernstein wave (EBW)
current drive potential, 402–403
heating, 362
Electron cyclotron current drive (ECCD), 23, 93–94, 109
Electron cyclotron emission (ECE) measurements, 571
Electron cyclotron heating (ECH), 571
Electron cyclotron resonance heating (ECRH), 472–473, 473f
Electron cyclotron (EC) systems
component layout and power supply circuit, 518, 519f
DIII-D system, 521–522
EUROfusion consortium, 522
gyrotron efficiency, 519–520
layout and plasma side, launch system, 518, 519f
power supply, 519–520, 520f
remote steering, 520, 521f
RF frequency, 520–521
short-pulse prototype gyrotron, 522
Energy storage system (EES), 80–81
Enhanced reversed shear (ERS), 131
Environmental constraints, magnetic fusion power plants
high plasma temperatures, 560–561
particle fluences, 564–565
radiation fluence, 563–564, 572f
radiation flux, 561–563, 561f
reflective first wall, 566
relativistic effects, 565
stray microwave absorption, 565–566
Ergodic divertor (ED), 282–283, 283f
European Fusion Development Agreement (EFDA), 216
European Power Plant Conceptual Study (PPCS), 580
Experimental advanced superconducting tokamak (EAST)
CFETR design, 432
DEMO divertor, 435
ELMy H-mode, 417–419, 418f
feeder system, 434, 434f
heating and current drive system, 409, 410f
high-performance long-pulse operation
divertor and SOL physics, 425–426, 426f
divertor power-deposition pattern, 427–428, 427f
edge and pedestal physics, 420–422, 421f
edge coherent mode (ECM), 415, 416f
edge-localized-modes (ELMs), 415, 422–425, 422f–425f
Greenwald density limit, 416
helical current, 415, 415f
H-mode discharge, 413–416, 414f, 417f
magnetic topology, 427
noninductive long-pulse H-mode plasma, 416–417, 418f
steady-state divertor power load, 428
steady-state long-pulse, 413–415, 413f
integrated advanced-operation scenario
beam torque and rotation-related quantities, 430–431, 431t
edge solutions, 431–432
ELMy H-mode, 429–430
RF systems, 430
steady-state plasma, 430
L- and H-mode discharges, 417–419, 419f
large superconducting magnet application, 435
long-pulse high-performance regimes, 411–412
machine, 409, 410f
1-MA discharge, 411–412, 411f
material and plasma evaluation system (MAPES), 428
mission and orientation, 412
monoblock mockups, 435
neoclassical toroidal plasma viscosity (NTV), 419, 420f
nonresonant magnetic braking studies, 420
parameters, 409, 410t
power supply, 433–434, 434f
resonant magnetic perturbation (RMP) coil system, 409–411
RF-dominated regimes, 417
steady-state core plasma operation, 428–429
superconducting conductors, 433, 433f
tungsten (W)-based materials, 435
F
Force-free helical reactor (FFHR)
See also Heliotron fusion power plant
DEMO designs, 583
FFHR-d1 type fusion-reactor, 585, 586f
helical divertor, advantage, 583
helical divertor coils, 584–585
heliotron reactor concept, 583
high-temperature superconductors (HTS), 584
large helical device (LHD), 581, 581f
low-temperature superconductor (LTS), 583–584
magnetic configuration, 582–583, 582f
magnetohydrodynamic (MHD) stability, 581
parameters, 585, 585t
reference blanket concept, 584
resonant magnetic perturbation (RMP), 584–585
Shafranov-shift, 582
Fuelling systems, 513
compact toroid (CT) technology, 540–542, 540f
edge-localized mode triggering, 541
effective fuelling rate, 535
flight systems, 542
gas puffing, 535
pellet injection technology, See Pellet injection technology
safety, 541
supersonic gas injection, 539–540
tritium pellets, 542
unmagnetized plasma jet, 541
Fusion blankets, 4
Fusion nuclear science facility (FNSF), 359
Fusion power plant (FPP), 61
G
Global energy economy, 3
Greenwald density, 11
H
Heating and current drive (HCD) system, 266–267, 268t
availability, 517
electron cyclotron (EC) systems, See Electron cyclotron (EC) systems
elements, 516
Fast Wave option, 535
high-power and long-pulse length requirements, 516
high power long pulse operations, 277–278
ion cyclotron resonance heating (ICRH) system, 279–280, 279f
ion cyclotron (IC) systems, See Ion cyclotron (IC) systems
lower hybrid current drive (LHCD) system, 278, 278f
lower hybrid (LH) systems, See Lower hybrid (LH) systems
neutral beam (NB) systems, See Neutral beam (NB) systems
NTM control, 510
phases, 509–510, 510f
plasma density and temperature, 516
ramp up phase, 510
reliability, 517
robust schemes, 280
suitability and limiting plasma factors, 509–510, 511t–512t
system code, 516
technology gaps, 534–535
Thomson Tubes Electroniques (THALES), 278
tritium breeding ratio, 517
wall plug efficiency, 517
Helical-axis advanced stellarator reactor
aspect ratio, 589
cooling system, 589
3-D neutronic investigations, 589
European fusion program, 588
Helias 5-B engineering study, 589, 591f
Helias reactors, 587–588
HTS technology, application, 588
magnetic configurations, 586–587
parameters, 589, 590t
Poincaré plot, 589, 590f
reactor-relevant criteria, 586–587
Wendelstein 7-X, 587, 587f
Heliotron fusion power plant
bridge-type lap joint, 489
FFHR-d1, 487–488, 488f
high-Tc superconductor (HTS), 489
parameters, 487, 488t
self-cooled liquid breeding blanket, 489
Helium-cooled pebble bed (HCPB), 65, 66f
I
Intermediate heat exchanger (IHX), 81
Intermediate heat transport and storage system (IHTS), 80–81
Ion cyclotron heating, 123–124
Ion cyclotron (IC) systems
coaxial transmission lines (TL), 524
comb array antenna, 524–525, 525f
disadvantages, 524
EUROfusion consortium, 526
Fibre Bragg Grating, 524
field aligned antenna, 525, 526f
layout, 523, 523f
near-term development program, 525
requirements, 522
RF radiation, 523–524
RF sheath, 525
J
Japanese and European fusion research communities, 442
Joint European Torus (JET)
carbon plasma-facing components, 215
disruptions
plasma current, centroid and poloidal magnetic field, 224–225, 225f
plasma-facing components, 227–228, 228f
prediction reliability, 228, 229f
quadratic scaling, 224–225
rate of disruptions, 226–227, 227f
thermal quench, 225–226, 226f
European Fusion Development Agreement (EFDA), 216
fusion diagnostics, 222–224, 223f–224f
fusion physics
Alfvén eigenmodes (AEs), 252
central electron temperature vs. alpha power, 250f, 251–252
classical Coulomb collisions, 250–251
deuterium–tritium mixtures, 253–254
diamagnetic stored energy, 253–254, 253f
gamma-ray spectra, 249f, 250–251
H-mode vs. the scaling, 251f, 253
normalised ion thermal conductivity vs. effective mass, 252f, 253
fusion reactor
active gas handling system (AGHS), 219
components, 216–217
deuterium–tritium experiment (DTE1), 218
deuterium–tritium experiment (DTE2), 219–220
edge localised mode (ELMs), 220–221
ion cyclotron resonance heating (ICRH), 220–221
JET ITER-like antenna, 220–221, 220f
lower hybrid current drive system, 219–220
mock-up tests, 217–218
parameters, 216, 217t
plasma conditions, 220–221, 221f
remote handling (RH) system, 216–218, 218f–219f
technical control system, 216
high-fusion performance
deuterium–tritium mixtures, 244–245
DTE1, 246, 248t
electron cyclotron current drive system, 247–248
highest fusion energy pulse, time history, 245–246, 246f
highest fusion power pulse, time history, 245–246, 245f
hot-ion H-modes, 245–246
Lawson criterion, 244–245
nonaxisymmetric magnetic fields, 247–248
noninductive current drive power, 247–248
optimised shear discharges, 246, 247f
tokamak plasmas, 247–248
JET Joint Undertaking (JJU), 216
plasma-wall interactions
beryllium divertor experiment, 231–232
divertor plasma geometry, 233
D retention rates, 232, 232f
exponential decay functions, 228–229, 230f
H-mode scrape-off layer, 229–230, 231f
ITER-like wall (ILW), 232
low-energy beryllium sputtering, 233
reactor-relevant conditions, 228
S-factor, 229–230
X-point configurations, 229–230
scrape-off layer
chamber neutral pressure, 233, 235f
gas-target solution, 235–237
ion-neutral collisions, 235–237
line-averaged density, 235–237, 236f
Mark IIA divertor, 233, 234f
radiated power fraction, 237, 237f
transport and confinement
dimensionless thermal energy confinement time vs. global-scaling relation, 239–240, 239f
high-performance hybrid discharges, 238–239
ion heat flux, 238–239, 238f
nonlinear electromagnetic turbulence, 238–239
normalised diffusive and convective transport, 241, 241f
tangential and normal injection, 240, 240f
turbulence mechanism, 238–239
JT-60SA
Broader Approach (BA) Satellite Tokamak Program, 439
burn simulation, high-β and high-bootstrap fraction, 462–463, 463f
characteristics
cryogenic systems, 446–447
cryostat, 446
divertor, 449–450, 450f
heating system, 447–449, 448f–449f
high-resolution diagnostics, 451
high-temperature superconducting current leads, 446
power supply systems, 447
stabilizing plates and in-vessel coils, 450–451, 450f
superconducting poloidal field coils, 445
superconducting toroidal field coils, 445
vacuum vessel, 446
components, 439–440, 441f
confinement and transport, 455–457, 457f
edge pedestal and edge-localized modes, 457–458
extended research phase, 465–467
high core plasma performances, 459
high-energy particle physics, 459
high-erosion and tritium-retention rates, 464–465
initial research phase, 465
integrated plasma performance, 453, 454f
integrated research phase, 465
ITER and DEMO
integrated plasma performance, 442, 443f
plasma control, 443–444, 444f
resistive wall mode (RWM), 444
self-regulating combined system, 442–443, 443f
theoretical models and simulation codes, 444–445
tritium breeding ratio (TBR), 444
Japanese and European fusion research communities, 442
MHD stability, 458–459
nondimensional parameters, 455, 456f
nuclear fusion energy, 439, 440f
operation scenarios, 454–455, 455f
plasma controllability, 460–462, 460f–462f
plasma equilibria, 451–453, 453f
plasma parameters, 451–453, 452t
radiative divertor, 459
replaceable divertor cassette, 463–464, 464f
research phases and components status, 465, 466t
safety factor profile, 454–455, 455f
SOL, 459
steady-state DEMO designs, 453–454, 454f
steady-state high-plasma-pressure operations, 439, 440f
steady-state safety factor profile, 459–460
test blanket modules, 463–464
test peripheral technology, 465
vacuum vessel, 440, 441f
JT-60U, 173f
active MHD stability control
neoclassical tearing mode suppression, 198, 199f
resistive wall mode suppression, 198–200, 200f
Advanced Material Tokamak Experiment (AMTEX) program, 168
DEMO Fusion Power Plant, 211
diagnostics, 174, 175f
disruption mitigation, 201–202, 202f
ECRF system, 174
edge transport barriers (ETBs), 174
energy confinement time, 169–170
ferritic steel tiles, 174, 176f
fusion plasma performance, 177–178, 178f
heat and particle control
Ar gas seeding, 204–206, 205f
He ash exhaust, 203, 204f
tungsten targets, 206–207, 207f
heating and current drive
high-resolution measurements, 167–168
high-βp experiments, 170
high-βp mode plasmas
central heating and beam fueling, 178
confinement improvement factor, 181, 181f
density and temperature profiles, 182
hot-ion plasmas, 178–180
internal transport barrier (ITB), 181, 182f
ion and electron temperatures, 180–182, 180f, 183f
parameters, 183–184, 186t
sawtooth-free target plasmas, 178–180
steady-state tokamak operation, 181
type-I ELMs, 182–183
βN vs. peaking factor, 182–183, 183f
H-mode confinement, 167
H-mode plasmas
grassy ELMs, 195, 195f
hybrid operation, 196–198, 197f
reduced fast ion loss, 195–196, 196f
threshold power, 194–195
hot-ion enhanced confinement, 170
hydrogen plasmas, 169
integrated plasma performance, 176, 177f
JT-60SA project, 211–212
negative-ion-based neutral beam (NNB), 168
parameters, 170, 172t, 177–178, 179t
perpendicular neutral beam (PNB) injection, 170–172
positive-ion-based neutral beam, 168
reversed shear (RS) mode plasmas, 167
current hole, 190–191, 191f
FULL code analyses, 188–189
high confinement and high beta, 192–193, 193f
high fBS discharge, 194, 194f
high βN value, 191–192, 192f
Motional Stark effect (MSE) measurement, 188
radial profiles, 188–189, 189f
RS configurations, 191–192
waveforms and profiles, 190, 190f
steady-state advanced tokamak operation, 176–177, 177f
superconducting tokamak, 168
tokamak physics experiment (TPX), 174
triangularity configurations, 172–173
vacuum vessel and the poloidal field coils, 170, 171f
W-shaped divertor, 168, 172–173, 173f
L
Large helical device (LHD)
characteristics
energy confinement, 477–479, 477f–478f
high-energy ions, 479
high-temperature magnetized plasma, 485
long-pulse operation, 484–485
magnetic axis, position, 476, 476f
MHD stability, 482
NB plasma initiation technique, 482
nondiffusive effect, 485
nonlinear coupling, 485
particle confinement, 480–482, 480f–481f
plasma edge physics, 483–484
plasma wall interaction, 484–485
quasi-steady state high-beta operation, 482, 483f
cryogenic system, 486
electron cyclotron resonance heating (ECRH), 472–473, 473f
heliotron concept, 469
heliotron fusion power plant, See Heliotron fusion power plant
ion cyclotron range of frequency heating, 474, 475f
local island divertor (LID), 471
magnetic coil system, 470–471, 470f
NB, plasma initiation, 475–476
negative ion, 486–487
neutral beam injection (NBI) heating, 474
parameters, 469, 469t
Poincare plot, 471–472, 471f
rotational transform, radial profile, 472, 472f
Wendelstein 7-X, 490, 490f
Large-scale clean energy, 3
Lawson criterion, 8–9, 244–245
Light water reactor technology, 81, 82f
Liquid breeder concepts, 66–67
Liquid metal breeder, 69
Lower hybrid (LH) systems
Frascati Tokamak Upgrade, 533
klystron amplifier, 532
launcher types, 532
layout and components, 530–532, 531f
Toshiba Electron Tube Devices, 532–533
M
Magnetic fusion power plants
burning phase, 513
diagnostics
blanket modules, 569
bolometry, 556–557
charge-exchange recombination spectroscopy (CER/CXRS), 555–556
contextual constraints, 560
control system, 550
DEMO and power plants, 569
device configuration, 559–560
diagnostic and port plug integration, 572
emerging data analysis techniques, 570–571
environmental constraints, 560
equilibrium control, 559
feedback-control loops, 550, 551t, 552
First Wall module, 571, 573f
heat and particle exhaust, 559
impurity content, 557
integration issues, 571
interferometry, 554
ITER toroidal interferometer and polarimeter (TIP), 571, 572f
magnetic equilibrium conditions, 570
measurement requirements, 557
motional stark effect (MSE), 553–554
optical components, 569
plasma current and density measurements, 559
polarimetry, 554
steady-state operation, 558
Thomson scattering, 554–555
total fusion power, 558–559
fuelling systems, 513
compact toroid (CT) technology, 540–542, 540f
edge-localized mode triggering, 541
effective fuelling rate, 535
flight systems, 542
gas puffing, 535
pellet injection technology, See Pellet injection technology
safety, 541
supersonic gas injection, 539–540
tritium pellets, 542
unmagnetized plasma jet, 541
HCDF systems, 513, 514t–515t
neutron irradiation, 518
requirements, 518
heating and current drive (HCD) system, See Heating and current drive (HCD) system
Lawson criteria triple product, 509
NTM instabilities, 513
requirements and capabilities, 509
specifications, 513
Magnetohydrodynamic (MHD) instabilities, 10–11
Magneto-hydrodynamic (MHD) pressure drops, 68–69
Massive-gas injection (MGI), 45
Material and plasma evaluation system (MAPES), 428
Medium-sized tokamak (MST), 93
Mega amp spherical tokamak (MAST)
aspect ratio, 359
CAD 3D configuration, 360, 360f
component test facility (CTF), 359
cross-field transport, 386
double null (DN) configuration, 386
D-shaped poloidal cross-section, 359
electron Bernstein wave (EBW) heating, 362
fast-ion physics and current drive
Alfvén instabilities, 382–383, 384f
MHD modes, 383–385
neutral beam current drive, 385, 386f
neutral beam ions, 382
self-sufficient burning plasma, 380
fusion nuclear science facility (FNSF), 359
heating system, 362
heat load distribution, 386
internal n = 1 kink mode
disruptions, 394–395
equilibrium reconstructions, 391
NTM physics, 393–394, 394f
sawtooth physics, 391–393, 392f
macroscopic stability, 390–395
magnetic confinement fusion research programme, 359
magnetic equilibrium, 360–361, 361f
MAST-U
Core scope, features, 399
divertor test facility, 400, 401f
facilities, 398
fully noninductive flat top, 402–403
fully noninductive start-up, 399
stages, 398–399
multibarrel pellet injector, 362
parameters, 360, 361t
plasma confinement
edge-localised modes, 363
energy confinement, 366–368, 367f
fuelling method, 369
high confinement (H-mode), 363–366, 365f
line-averaged density, 368, 369f
momentum transport, 370, 371f
particle confinement, 368
pellet retention time, 368, 369f
plasma start-up
ECRH/EBW start-up and heating, 396–397, 397f
HFS coils, neutron shielding, 395
merging compression start-up, 395–396
scrape-off layer (SOL) transport, 388–389, 388f, 390f
Small Tight Aspect Ratio Tokamak (START), 359
spatiotemporal resolution, 370
target heat loads, 387–388
turbulent core transport, 370
Momentum exhaust
heat flux profile, 39, 40f
ion–electron pairs, 38
momentum loss factor, 39, 39f
momentum ratio, 38–39
multifaceted axisymmetric radiation, 40
radiated power fraction vs. line-averaged density, 40–41, 42f
time evolution, 40, 41f
N
National Spherical Torus eXperiment (NSTX) Upgrade
bootstrap and beam-driven noninductive, 352
boundary physics
cross-field transport, 341
3D heat-conduction solver, 344
edge and SOL turbulence measurements, 344–345
ITER-class devices, 346–347
light emission, GPI images, 344–345, 344f
liquid lithium modules, 345–346
midplane scrape-off layer widths, 342–344, 343f
poloidal and heat flux, 345, 346f
poloidal Reynolds acceleration, 344–345, 345f
Princeton Plasma Physics Laboratory, 347
snow flake divertor (SFD), 345
temporal edge, 342, 343f
type I ELMs, 342
unlithiated and lithiated discharges, 341–342, 342f
diagnostics, 326–327
energetic particles
Alfvén activity, types, 337–338, 339f
Alfvén velocity, 336–337
EP-driven MHD modes, 341
EP-induced magnetic fluctuations, 337, 338f
fishbone-like energetic particle mode, 338–340
high-frequency modes, 337
mode stability, 338–340, 339f
nonresonant kink (NRK), 338–340
parallel magnetic field perturbation and electric field contours, 340–341, 340f
toroidal Alfvén eigenmodes (TAEs), 337
ETG modes, 352
fully noninductive current, 354–355, 354f
high-flux expansion snow flake/X-divertors, 353–354
interior view, 326, 326f
macroscopic stability
disruptivity, 335, 336f
error fields (EFs), 334
MHD stability database, 332–333, 333f
multi-input criteria, 335–336
neoclassical tearing modes (NTMs), 334
noninductive bootstrap fraction, 332
resonant field amplification (RFA), 334, 335f
RWM amplitude, 334
time evolution, 332–333, 333f
warning times, 335–336, 337f
Magnum-PSI divertor test stand, 353–354
mission element, 325
normalized confinement time, 351–352, 351f–352f
vs. NSTX, 352, 353f
plasma-facing component (PFC), 326
research goals, 350–351
solenoid-free operation and wave physics, 347–350, 348f, 350f–351f
transport and turbulence
CAE/GAE activity, 331
collisionality and electron, 329–330, 330f
electron and ion thermal, 329–330, 330f
electron-gyroradius scales, 331
energy diffusivity profiles, 329–330, 330f
H-mode discharges, 329–330
ion thermal diffusivity, 331, 332f
L-mode global confinement scaling, 328–329
microtearing modes, 330–331
normalized thermal confinement time, 328–329, 329f
thermal energy confinement scaling, 328–329, 328f
toroidicity, 327–328
wall conditioning, 353
Negative-ion-based neutral beam (NNB), 168
Neoclassical tearing modes (NTMs), 11, 21–22, 145–146, 334
Neoclassical toroidal plasma viscosity (NTV), 419, 420f
Neutral beam current drive (NBCD) system, 402
Neutral beam injection (NBI) heating, 123, 207, 474
Neutral beam (NB) systems
actively cooled duct, 528
beamline components, 528
breeding blanket, 528–529
Broader Approach agreement, 530
efficiency, 529
EUROfusion consortium, 530
multiple beam system and ITER 1-MV system, 526, 527f
negative ion-based systems, 529
negative ion sources, 529–530
photoneutralizer, 530, 531f
power absorption, 528
reliability and availability, 529
SIPHORE project, 530
unneutralized ions, 528
P
Pellet injection technology
anticipated fuelling rates, 538–539
differential pumping system, 538
elements, 536, 536f
ELM pacing pellet, 538
extrusion rate, 538
maximum production rate, 537–538
pellet acceleration methods, 537
pellet mass loss, 537
steady-state operation, 538
volume and speed distribution, 538, 539f
Plasma exhaust
divertor heat exhaust channel, 53–55, 54f
energy confinement, metal wall, 52–53, 52f–53f
exhaust requirements, 32–33
innovative divertor schemes, 56–57, 56f
ITER tokamak, poloidal cross-section, 32, 33f
magnetic topology, 32
momentum exhaust, See Momentum exhaust
particle exhaust, 33–34, 35f
plasma–material interactions (PMI), 31, 32f
blobs/intermittent plasma objects, 46–47
He–I light emission, 46–47, 47f
main chamber recycling, 47
plasma facing materials, See Plasma facing materials
SOL density, 47, 48f
power exhaust, 38
See also Power exhaust
steady-state peak heat flux, 55
thermonuclear fusion, 31
transients and 3D effects
current quench (CQ), 45
disruptions, 44–45
edge localized modes, 41–44, 43f–44f
peak heat flux, 45, 45f
plasma/magnetic field stored energy, 41
power-producing reactors, 45
shattered pellet and massive-gas injection, 45, 46f
substantial toroidal asymmetries, 45–46
thermal quench (TQ), 45
Plasma facing materials, 4
liquids, 50–51, 51f
solids
boronization, 49–50
carbon, 47–48
fuzz formation, 49, 50f
gas fueling, 48–49
ion cyclotron radio frequency (ICRF), 48–49
tungsten, 48
Plasma performance, burn and sustainment
Alfvén Eigenmodes, 12
bootstrap current density, 13
bootstrap fraction, 13
energy confinement time, 8–9
external heating power, 12
fusion reactions, 8–9
fusion reactor, requirements, 14–16, 14f–15f
α-heating, 23–25, 24f
ITER vs. DEMO, 29, 29t
Lawson criterion, 8–9
nested magnetic surfaces, 7
plasma operation scenario, 13
poloidal beta, 13
safety factor, 8
self-heating, α-particles, 12
stability
density limit, 11
edge localised mode (ELM), 21
electron cyclotron current drive (ECCD), 23
Greenwald density, 11
helical perturbation magnetic fields, 20–21, 20f
magnetohydrodynamic (MHD) instabilities, 10–11
Mirnov activity, 22, 22f
neoclassical tearing modes (NTMs), 11, 21–22
quasi-stationary ELMing H-mode discharge, 21, 21f
resistive wall mode (RWM), 20–21
tearing modes, 11
Troyon-limit, 10–11
Troyon scaling, 20
tokamak operational scenarios
advanced operational scenarios, 25
bootstrap fraction, 27
conventional scenarios, 25
core radiation, 28
DEMO/fusion reactor, 26
L-mode, 26
noninductive operation, 27–28, 28f
power threshold, 26
pressure and safety factor, 25, 26f
reversed shear scenario, 27
toroidal magnetic geometry, 7, 7f
toroidicity-induced Alfvén eigenmodes (TAEs), 12
transformer flux, 13
transport
ASDEX Upgrade tokamak, 17–18, 18f
cross-field transport, 9
H-factor, 9–10
H-mode conditions, 19
linear stability, 17, 17f
quasi-linear approach, 16–17
refined collisional transport, 9
self-consistent ab-initio turbulence simulation, 19–20
transport barriers, 18–19, 19f
turbulent transport, 9–10, 16
Power exhaust
central line-averaged density, 106–108
charge exchange process, 37
cross-field transport, 36
differential phase angle, 106–108, 108f
dissipative process, 34–35
divertor dissipation, 37
dominant dissipative process, 37, 38f
heat flux, 36
magnetic flux expansion ratio, 36
magnetic perturbation coils, 105–106, 106f
neoclassical toroidal viscosity, 105–106
peak divertor heat flux, 36
plasma density, 106–108
power decay length, 105
private flux region (PFR), 37
steady-state power balance, 34–35
W influx and W concentration, 106, 107f
Power extraction
balance of plant (BOP), See Balance of plant (BOP)
components
breeding blanket, 76–77
ITER divertor, W-mono-block confi guration, 74–76, 75f
neutron first wall loading, 71, 72f
neutron irradiation, 76
neutron wall loading, 71
PbLi flow channel, 73
plasma heat flux, 74, 75f
power density, 72–73
radial volumetric heat power distribution, 72, 73f
radiative flow distribution, 73, 74f
thermal conductivity, 72–73
thermo-mechanic performances, 76
primary heat transfer system (PHTS), 77–79, 77f–78f
Pressure water reactor (PWR), 65–66
Primary heat transfer system (PHTS), 77–79, 77f–78f
R
Rankine system, 70–71
Remote handling (RH) system, 216–218, 218f–219f
Resistive wall mode (RWM), 20–21
Resonant magnetic perturbation (RMP) coil system, 409–411
Reversed shear (RS) mode, 131, 167
current hole, 190–191, 191f
FULL code analyses, 188–189
high confinement and high beta, 192–193, 193f
high fBS discharge, 194, 194f
high βN value, 191–192, 192f
Motional Stark effect (MSE) measurement, 188
radial profiles, 188–189, 189f
RS configurations, 191–192
waveforms and profiles, 190, 190f
Rotating electrode method, 65
S
Scrape-off layer (SOL) transport, 388–389, 388f, 390f
Shattered pellet injection (SPI), 45
Small Tight Aspect Ratio Tokamak (START), 359
Snow flake divertor (SFD), 345
Solid breeder blankets, 64–65, 65f
Solid breeder technology, 64
Steam turbines, 79, 80f
Stellarator fusion power plants
advantages, 577–578
Alfvénic instabilities, 596
blanket modules, 579
compact stellarator
ARIES-CS power plant, See ARIES-CS power plant
construction costs, 591
Dual Coolant Lithium Lead (DCLL) configuration, 594
Nb3Sn technology, 593
parameters, 595, 595t
particle exhaust, 594–595
QAS configurations, 591–592
tungsten–carbide (WC) shield, 594
cross-fertilization, 595
European Power Plant Conceptual Study (PPCS), 580
force-free helical reactor (FFHR), See Force-free helical reactor (FFHR)
helical-axis advanced stellarator reactor, See Helical-axis advanced stellarator reactor
helical devices, 577
maintenance period, 579–580
Model-C stellarator, 578
neoclassical tearing modes, 577
quasiaxisymmetric compact stellarator concept, 597
Shafranov shift, 596
support-structure scales, 580
toroidal current, 577
tritium breeding ratio (TBR), 578–579
Superconducting Magnet Power Supply (SCMPS), 447
T
TBM and EU DEMO blanket, 69–70, 70t
Test Blanket Programme, 69–70
Thermal power, 61
Tokamak device, 4
Tokamak fusion test reactor (TFTR)
activated/contaminated components, 127
bootstrap current, 141
boronization, 122–123
bumper limiter, 121–122
carbon blooms, 121–122
confinement zone, 161
cryogenic distillation system, 125
current, 143
density, 143, 144f
design parameters, 119, 120t
diagnostics, 127, 128t
disruptions and avoidance, 146–147
D-T experiments, 126
alpha-driven instabilities, 159, 160f
alpha heating, 157–158, 158f
alpha loss rate vs. pitch angle and gyroradius, 152, 153f
alpha particles, 152, 153f
α-CHERS signals, 155, 156f
deuterium–tritium beam injection, 156–157, 157f
energy confinement, 148–151, 150f–151f
fast-ion confinement, 151–152
fusion power optimization, 148, 149f
helium ash, 156–157
lost-alpha detectors, 153
NBI sources, 148
non-sawtoothing D-T supershot, 155, 156f
particle confinement, 151, 152f
PCX measurements, 154–155, 154f–155f
plasma supershots, 155
supra-thermal alpha particles, 155
fast-ion confinement, 141
fueling and impurity injection systems, 124
heating systems
ion cyclotron heating, 123–124
neutral beam injection (NBI), 123
ICRF physics, 141–143
jaws limiter, 121
kink/ballooning mode, 144–145, 145f
magnets, 121
neoclassical resistivity, 141
neoclassical tearing modes (NTMs), 145–146
neutral beam injection–heated plasmas
detached plasmas, 131–132
enhanced reversed shear (ERS), 131
high internal inductance regime, 130–131
H-mode, 130
L-mode, 129
radiating mantles, 131–132
reversed shear (RS), 131
supershots, 129–130
neutron irradiation, 126–127
normalized beta, 144
ohmically heated plasmas, 128–129
plasma composition, 132
plasma-facing components (PFCs), 122
pressure-driven MHD instabilities, 143–144
pulse discharge cleaning (PDC), 122
shielding, 126
test cell, 119, 120f
thermal energy and particle confinement
back transition, 137–139, 139f
beam emission spectroscopy (BES), 136–137, 137f
constant-current discharges, 135–136
core confinement region, 139–140
deuterium plasmas, 132
effective ion thermal diffusivity, 134, 135f
ERS transition, 137–139, 139f
fluctuation-driven transport, 136
gyro-Bohm scaling, 133
heat pulse propagation, 140
H-mode confinement, 135
ion temperature gradient (ITG) mode, 134–135
L-mode plasmas, 133
nondimensional plasma parameters, 133
ohmically heated plasmas, 132
perturbative transport studies, 140
spatial correlation functions and wave number spectra, 137, 138f
supershot confinement, 133–134
thermal plasma confinement, 133
transport coefficients, 134
turbulent fluctuations, 136
volume-average density, 134
thermalization, 141
tritium gas delivery and exhaust system, 124, 125f
tritium retention and removal, 125–126
US Atomic Energy Commission, 119
vacuum vessel, 121–122, 122f
wall interactions, 146
Tore Supra
actively cooled endoscopes, 269, 270f
actively cooled plasma facing components
actively cooled cooper alloy, 277
brazed first wall concept, 273, 275f
characteristics, 273, 274t
configuration, 277, 277f
heat exhaust, 275
heat flux, 276
plasma–wall interaction, 277
thermal and mechanical properties, 275–276
toroidal pump limiter, 276–277, 276f
AIA robot, 269–270, 270f
assembly, 265, 267f
centrifugal pellet injector, 268–269
characteristics, 263, 264t
CIEL, 262
components, 265, 266f
diagnostics system, 269
ergodic divertor (ED) experiments, 282–283, 283f
European fusion roadmap, 261
Frank–Condon neutrals, 269
French Tokamak Tore Supra, 261
fueling systems, 267–268, 268t
heating and current drive systems, 266–267, 268t
high power long pulse operations, 277–278
ion cyclotron resonance heating (ICRH) system, 279–280, 279f
lower hybrid current drive (LHCD) system, 278, 278f
robust schemes, 280
Thomson Tubes Electroniques (THALES), 278
inner vacuum vessel, 263, 266f
IR thermography system, 269
ITER operational issues, 280–282, 281f
long pulse experiments
accumulated carbon deposits, 284–285
constant retentions rate, 286
cumulated gas injection and wall inventory, 286, 286f
densities and non-inductive current, 285
LHEP regime, 285–286
plasma parameters, 284, 284f
stable plasma scenario, 284
steady-state plasma control, 283
time constant, 284–285
toroidal pump limiter sector, 284–285, 285f
long pulse operation, 265–266
lower hybrid current drive system, 266–267
meridian cross-section, 263, 264t
monitoring systems, 271
niobium titanium (Nb–Ti), 262
plasma–wall equilibrium time-scale, 291, 291f
superconducting magnet, 262–263, 271–273, 271t, 272f
thermal losses, 263
toroidal magnet, 263
Tungsten (W) Environment in Steady-State Tokamak (WEST) project, 292
carbon to tungsten PFC, 288–289, 289f–290f
ITER divertor PFC, 289–290
ITER plans, 287
limiter to divertor configuration, 288, 288f
long pulse H-mode operation, 290–291
water-cooled copper coils, 262
Toroidal Alfvén eigenmodes (TAEs), 337
Toroidicity-induced Alfvén eigenmodes (TAEs), 12
Tritium production
fuel cycle, 83–84, 83f
HCLL system and T control mechanisms, 86, 86f
helium purge-gas transports, 86
liquid breeder blankets, 85
neutron multiplication, 85
thermonuclear reaction, 84
trapping mechanisms, 84
tritium breeding ratio (TBR), 84
tritium residence time (TRT), 85
V-Li blanket concept, 84
U
US ARIES Programme, 68–69
US Atomic Energy Commission, 119
W
Watershed metaphor, 4–5
WCCB Blanket concept, 65–66, 67f
WCLL blanket, 67–68, 68f
Wendelstein 7-X, 490, 490f
assembly, 501–502, 502f
cold mass, 498
completed stellarator device, 503, 503f
construction, 497, 498f
10 divertor modules, 500–501, 500f
initial research phases, 503–504, 504f
in-vessel components, 499–500, 500f
last closed flux surface (LCFS), 496
magnetic field geometry optimization, 494
magnet module, 497–498
modular coil set, 587, 587f
nonplanar coils, 495, 495f
optimization procedure, 496, 497f
parameters, 497, 499t
physics-optimized magnetic field geometry, 494
radial pressure gradient, 496–497
requirements, 494–495
stellarator family, 493
stellarator optimization, 496
thermal insulation, 499
three-dimensional geometry, 501–502
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